Separate Effect Test Facility Research Articles

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

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

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

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

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

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

Computational model development and validation of fuel dispersal phenomena

Computational model development and validation of fuel dispersal phenomena

The LIFUS5 separate effect test facility experimental programme

The LIFUS5 separate effect test facility experimental programme

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.

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

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

Modeling of horizontal steam injection in a water pool

Modeling of horizontal steam injection in a water pool

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

• Obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. • The RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time. • The CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model. 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.

Review and outlook of the integral test facility PKL III corresponding studies

Abstract This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.

Steam condensation separate effect tests in core make-up tank

The core make-up tank (CMT) is a crucial device of the passive core cooling system in the Westinghouse AP-series pressurized water reactor nuclear power plants . In the prototype plant, the CMT is a vertical cylindrical vessel with hemispherical top and bottom end sockets. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors affect the performance of the passive core cooling system (PXS). Therefore, it is necessary to carry out separate effect tests to verify the feasibility of CMT applied to passive safety and injection systems and to study the operating characteristics of CMT. This article first presents the design of a CMT separate effect test facility and analysis of the operating characteristics of the CMT with the experimental data. The influence of pressure, non-condensable gas, wall condensation, and thermal resistance on the CMT behavior were investigated. The study found that the non-condensable gas would stratify in the CMT, which had negligible effect on the wall condensation, but could greatly reduce the gas-liquid interface condensation.

Test Series D experimental results for SIMMER code validation of WCLL BB in-box LOCA in LIFUS5/Mod3 facility

The in-box LOCA (Loss of Coolant Accident) is a relevant safety issue for the design of the WCLL BB. Research activities are ongoing to master phenomena and processes occurring during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes and procedures for their applications. Current status of knowledge requires the availability of qualified and reliable experimental data to support these activities. In view of this, the new separate effect test facility LIFUS5/Mod3 has been commissioned and the Series D experimental campaign is in progress. The tests are focused on the generation of reliable experimental data for the validation of the modified version SIMMER codes for fusion application. Moreover, data will be also used to investigate the dynamic effects of energy release on the structures and to provide relevant feedbacks for the follow up experimental campaigns. The experimental data and results of Test D1.1 are reported and critically discussed, focusing on pressures, temperatures, amount of injected water, and hydrogen production quantification.

An overview on thermal-hydraulic phenomena for water cooled nuclear reactors; part I: SETs, and ITFs of PWRs, BWRs, VVERs

Abstract Validation of best-estimate thermal-hydraulic system codes is a necessary step to prove their applicability to calculate nuclear plant accident scenarios, including the course of events. It shall demonstrate that those physical thermal-hydraulic phenomena, which are important for a nuclear plant transient and accident scenarios, are calculated and analysed appropriately for the design and safety analysis purposes. Within the framework of international organizations, e. g., the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of The Organization for Economic Cooperation and Development (OECD/NEA-CSNI) and the International Atomic Energy Agency (IAEA), with the major support of institutions from different countries, work has been carried out for establishing validation matrices on Separate Effects Test (SET) facilities and Integral Test facilities of PWRs and BWRs. During this process, one of the major elements of the activity was to identify and characterize the thermal-hydraulic phenomena and, also test types. This process was based on expert knowledge of the scientists involved and, approved by the participating organizations and institutions in different countries. Additionally, identification and characterization of thermal-hydraulic phenomena were performed for VVERs, ALWRs (including SMRs) and, also SCWRs with emphasize to the design specifics of these nuclear plants. A common list of 67 thermal-hydraulic phenomena including basic phenomena is provided for Separate Effects Tests (SETs), and Integral Test facilities (ITFs) of PWRs, BWRs, and VVERs have about 50 thermal-hydraulic phenomena. Additionally, Advanced Water-Cooled Reactors having 15 identified phenomena, as well as SCWRs (specifically HPLWR) related 6 more phenomena identified. For all these internationally agreed thermal-hydraulic phenomena either some detailed descriptions for selected cases are given or needed references are provided in the paper. These identified phenomena also help to define research needs in the related field for the major number of water-cooled reactors. The knowledge of the thermal-hydraulic phenomena is the backbone and essential part for the development of nuclear thermal-hydraulics subject area. Thus, the subject will be presented in two parts: Part I of the paper will be devoted to thermal-hydraulic phenomena related to SETs, PWRs, BWRs, and VVER-400 plus VVER-1000; Part II will be dealing with additional thermal-hydraulic phenomena for ALWRs including VVER-1200 and SMRs, and SCWRs which are covered under Gen-III+ and GEN-IV designs.

An overview on thermal-hydraulic phenomena for water cooled nuclear reactors; part II: ALWRs and SCWRs

Validation of best-estimate thermal-hydraulic system codes is a necessary step to prove their applicability to calculate nuclear plant accident scenarios, including the course of events. It shall demonstrate that those physical thermal-hydraulic phenomena, which are important for a nuclear plant transient and accident scenarios, are calculated and analysed appropriately for the design and safety analysis purposes. Within the framework of international organizations, e. g., the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of The Organization for Economic Cooperation and Development (OECD/NEA-CSNI) and the International Atomic Energy Agency (IAEA), with the major support of institutions from different countries, work has been carried out for establishing validation matrices on Separate Effects Test (SET) facilities and Integral Test facilities of PWRs and BWRs. During this process, one of the major elements of the activity was to identify and characterize the thermal-hydraulic phenomena and, also test types. This process was based on expert knowledge of the scientists involved and, approved by the participating organizations and institutions in different countries. Additionally, identification and characterization of thermal-hydraulic phenomena were performed for VVERs, ALWRs (including SMRs) and, also SCWRs with emphasize to the design specifics of these nuclear plants. A common list of 67 thermal-hydraulic phenomena including basic phenomena is provided for Separate Effects Tests (SETs), and Integral Test facilities (ITFs) of PWRs, BWRs, and VVERs have about 50 thermal-hydraulic phenomena. Additionally, Advanced Water-Cooled Reactors having 15 identified phenomena, as well as SCWRs (specifically HPLWR) related 7 more phenomena identified. For all these internationally agreed thermal-hydraulic phenomena either some detailed descriptions for selected cases are given or needed references are provided in the paper. These identified phenomena also help to define research needs in the related field for the major number of water-cooled reactors. The knowledge of the thermal-hydraulic phenomena is the backbone and essential part for the development of nuclear thermal-hydraulics subject area. Thus, the subject will be presented in two parts: Part I of the paper will be devoted to thermal-hydraulic phenomena related to SETs, PWRs, BWRs, and VVER-400 plus VVER-1000; Part II will be dealing with additional thermal-hydraulic phenomena for ALWRs including VVER-1200, and SCWRs which are covered under GEN-III+ and GEN-IV designs.

An experimental study on the thermal-hydraulic phenomena in the Hybrid Safety Injection Tank using a separate effect test facility

Abstract This paper reports an experimental research for investigating thermal hydraulic phenomena of Hybrid Safety Injection Tank (Hybrid SIT) using a separate effect test facility in Korea Atomic Energy Research Institute (KAERI). The Hybrid SIT is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer (PZR). The separate effect test (SET) facility of Hybrid SIT, which is designed based on the APR+ power plant, comprises a PZR, Hybrid SIT, pressure balancing line (PBL), injection line (IL), nitrogen gas line, and refueling water tank (RWT). Furthermore, the pressure loss range of the SET facility was analyzed and compared with that of the reference nuclear power plant. In this research, a condition for balancing the pressure between the Hybrid SIT and PZR is examined and the effects of different variables affecting the pressure balancing, which are flow rate, injection velocity of steam and initial water level, are also investigated. The condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel. Additionally, a sensitivity analysis as a theoretical approach was conducted on the pressure variations in relation to the rate of steam condensation inside the Hybrid SIT. The results showed that pressure of the Hybrid SIT was predominantly determined by the rate of steam condensation. The results showed that if the rate of condensation increased or decreased by 10%, the Hybrid SIT pressure at the pressure balancing point decreased or increased roughly by 10%, respectively.

Addressing the heavy liquid metal – Water interaction issue in LBE system

Abstract The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.

Heat Transfer Characteristics of Passive Condensers for Reactor Containment Cooling

Passive condensers that are based on gravitational force do not require pumps or blowers to move fluid and, hence, are considered as safety measures in nuclear power plant containment cooling. In this paper vertical tube passive condensers with the tube inside condensation and cooled by a secondary pool of water are discussed as applied to reactor containment heat removal system. Series of experiments were carried out on scaled separate effects test facilities with vertical single tube and four tube bundle condensers. The results of experimental data on the rate of condensation heat transfer, the effects of noncondensable, and parametric effects, such as pressure, tube diameters and lengths, obtained from single tube and tube bundle passive condenser are presented. The data on the tube bundle indicated larger condensation heat transfer due to enhanced heat transfer in the secondary cooling. For tube bundle condensation, the turbulent mixing on the secondary pool side decreases the temperature difference between pool water and condenser tube outer wall, causing an increase in secondary heat transfer. Theoretical models on tube condensation based on heat and mass analogy and boundary layer models were developed. The data were compared with theoretical models and good agreement was observed between theoretical predictions and single tube condensation heat transfer data. For the tube bundle condenser, the theoretical model agreed with data when the secondary side enhancement in heat transfer due to turbulent mixing in the flow boiling was included. Practical heat transfer correlations are presented using available data on passive condensers for reactor application.

An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger

Abstract The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance.