Assessment of SBO Fukushima likewise scenario for an IPWR design with RELAP5MOD33 and RELAPSCDAPSIMMOD3. 5 codes

In recent years Small Modular Reactors (SMR) have become very popular within the nuclear industry. These designs allow to reduce costs as well as to enhance the safety due to passive nuclear safety features. Within these systems, the integral Pressurized Water Reactors (iPWR) are very extended because they take advantage of the previous technology developed for Gen II and III PWRs. In this sense, previous Best Estimate system codes like RELAP5 or CATHARE seem to be reliable for Deterministic Safety Assessment (DSA) but need to be assessed for new passive systems in which natural circulation takes a key role. In the present paper, Energy Software Ltd., in collaboration with the UPC, has developed an iPWR input model for both NRC RELAP5 and ISS RELAPSCDAPSIM codes. These models, based on CAREM-25 publicly available data, simulate an SBO Fukushima likewise scenario. Results under Design Basis Accident (DBA) conditions are benchmarked to assess the reliability of the codes to reproduce the plant availability reported in the collected data. Passive systems like Safety Injections and Residual Heat Removal Exchangers have also been included to analyze the code capabilities to reproduce natural circulation under iPWR conditions. Finally, core damage progression is simulated with SCDAP components to analyze the severe accident related phenomena. Results of both simulations seem to confirm the 36 hours grace period for SBO scenario of the CAREM-25 design plus the extended 36 hours grace period associated to the availability of Emergency Injection System (EIS) in Loss of Coolant conditions reported by designer.Peer ReviewedPostprint (published version

Methodology for phenomenological code assessment with integral test data

The use of codes in the licensing process requires a rigorous validation process that can be accomplished by means of qualitative and quantitative assessment. In thermal hydraulics, this validation has to be performed at different levels, from separate effects to the integral response of a plant design. Even though the quantitative assessment is preferred, for complex phenomenology involving the behaviour of the whole plant system this approach is difficult and the assessment is usually performed through qualitative expert judgement. In the present article, a methodology is proposed that combines the use of qualitative and quantitative adequacy assessment for the simulation of experiments at integral test facilities. The method makes use of statistical quantification by means of Best Estimate Plus Uncertainty calculations.Peer ReviewedPostprint (published version

Application of a BEPU-based code assessment to the ATLAS upper head SB-LOCA test

The prevailing state of knowledge of two-phase flow in complex systems, as in the nuclear field, usually leads a certain degree of ad hoc calibration of computational models and the consequent inevitable subjectivity in analyses and assessments. The UPC-ANT Uncertainty Analysis methodology for code assessment attempts to maintain scientifically desired features of models and analyses, by crediting BEPU analysis and centering the assessment criteria on T-H phenomenology rather than on safety criteria, as it is the usual practice. An SB-LOCA in the upper head is utilized as a case study, and global, chronological, FOMs and phenomenological analyses are carried out and illustrated briefly. Overall, it was concluded that the developed RELAP5 model reproduces all the relevant T-H phenomena to the investigated scenario, either partially or totally, and the BEPU methodology is generally adequate to perturb the underlying T-H phenomena.Peer ReviewedObjectius de Desenvolupament Sostenible::7 - Energia Assequible i No ContaminantObjectius de Desenvolupament Sostenible::9 - Indústria, Innovació i InfraestructuraPostprint (published version

Revisiting ISP-13 with RELAP/SCDAPSIM/MOD3.5 using core SCDAP components

The recent accident in the Fukushima Daiichi nuclear power plant opened a discussion on severe accident management that includes the analysis of the accident by means of computational tools that can predict the core behavior in such extreme conditions. The RELAP/SCDAPSIM/MOD3.5 code is designed to predict the behavior of Light Water Reactor (LWR) coolant systems during normal and accident conditions including severe accidents up to the point of reactor vessel failure. The code consists of two parts: the RELAP5 models calculate the overall Reactor Coolant System (RCS) thermal-hydraulic response, control system behavior, reactor kinetics and the behavior of special reactor system components such as valves and pumps, to predict the plant behavior under operational transients, Design Basis Accidents (DBAs) and Beyond DBAs; the SCDAP models calculate the behavior of the core and vessel structures under normal and severe accident conditions. Both portions of the code have been proven, separately, to accurately reproduce the response under its designed purpose, which are steady state, DBAs and BDBAs for the RELAP portion, and steady state and severe accident conditions for the SCDAP portion. The analysis of potential scenarios does not define a priori the final state of the fuel rods, and thus the most adequate tool is a system code such as RELAP/SCDAPSIM/MOD3.5 capable of simulating accident scenarios where severe accident phenomena may or may not occur. The present paper revisits the ISP-13 exercise, a cold leg double-ended guillotine LOCA conducted in the LOFT experimental facility, using two RELAP/SCDAPSIM/MOD3.5 models: the first one is entirely modeled with RELAP components, the second model keeps the RELAP nodalization with the exception of the core region, which is modeled with SCDAP components. The LOFT L2.5 experiment is a rather unique experiment since it features nuclear (UO2) fuel rods in a facility designed to simulate the major responses of a commercial pressurized water reactor (PWR). In addition, the fuel cladding of this experiment reached relatively high temperatures of around 1100 K. Even though this cladding temperature is far from the oxidation onset with steam, the LOFT L2-5 experiment challenges system behavior simulations by bringing the conditions close to those of severe accidents. The final goal is to evaluate whether the use of SCDAP components in LOFT L2-5 experiment reproduces similar results to those obtained with a RELAP standalone model, and that both simulations are in good agreement with experimental data.Postprint (published version

RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of test blanket modules involving helium flows into heavy liquid metal

The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. Thermal hydraulic safety analyses are being carried out with the system code RELAP/SCDAPSIM/MOD4.0 whic h was initially designed to predict the behaviour of light water reactor systems during normal and accident conditions. The code is being developed as part of the international SCDAP Development and Training Pr ogram (SDTP) coordinated by Innovative Systems Software (ISS). The modeling strategy of the RELAP code for the simulation of two-phase flows is based on a single-fluid two-phase approach with a set of momentum, energy and mass equations for each phase. The two phases are liquid-water and gas phase mixture of steam and non-condensable gases. Phase interactions, such as interphase friction and heat transfer, are modelled by closure relations based on experimental/numerical correlations that depend on the flow regime. In cooperation with ISS, the IPR team has implemented LLE liquid phase thermodynamic properties as a working fluid alternative to water and appropriate wall-to-LLE heat transfer correlations. However, in order to analyze some of the postulated off-normal events, there is a need to simulate the mixing of helium and Lead-Lithium fluids In the standard RELAP/SCDAPSIM/MOD4.0 ve rsion it is not possible to simulate a mixture of a non-water fluid with a non-condensable. In addition to that, t he available flow regime maps for vertical and horizontal flows in RELAP are specific for steam/water pair, which may not be suitable for LLE/helium pair. The Technical University of Catalonia is cooperating with IPR to adapt the RELAP/SCDAPSIM/MOD4.0 code to allow the si mulation of LLE and he mixture. This paper presents the results of the first step of the project, which includes a state of the art on simulation of liquid metals mixed with non-condensable using system codes, the implementation of the necessary code modifications to allow for a LLE/he mixture a nd preliminary results using the modified code version for horizontal and vertical configurations.Postprint (published version

Linden code validation against NUPEC / PSBT experimental data for void fraction and temperature benchmarks

The subchannel analysis code LINDEN is being developed by China Nuclear Power Technology Research Institute Co. Ltd (CNPRI). The LINDEN code is used in thermal-hydraulics design and safety analysis of pressurized water reactor (PWR) cores. As part of the code development activities, CNPRI commissioned Energy Software Ltd. (ENSO) to conduct an independent assessment of the LINDEN code. The experimental data from Nuclear Power Engineering Corporation (NUPEC) PWR Subchannel and Bundle Tests (PSBT), available through the PSBT benchmark activity, were selected for this validation. The assessment work focused on the void fraction and temperature related benchmarks and was divided into three parts: (1) steady-state void fraction and pressure drop benchmarks, (2) steady-state f luid temperature benchmark, and (3) transient void fraction benchmark. The results presented in this paper correspond to the steady-state parts of the validation work. The assessment of the code comprised the code-to-data comparison as well as the code-to-code comparison. The first one relied on the concepts of accuracy, precision and consistency which can be quantitatively evaluated from statistical indicators and their comparison to the uncertainty of the experimental measurements. The second one consisted in a qualitative assessment against the PSBT benchmark results, with the object of comparing the LINDEN calculated values to other state-of-the-art codes for this type of analysis, and of complementing the code-to-data comparison, which lacked precise information on the experimental uncertainty. The overall conclusion is that the LINDEN calculated values can be considered in good agreement with the PSBT data.Peer ReviewedPostprint (published version

Significance of the input parameters selection and the nodalization qualification in the final results of an IBLOCA BEPU calculation

In the framework of Design Basis Accidents (DBA) for Pressurized Water Reactor (PWR), and based on recent studies on pipe integrity combined with the Risk-Informed Decision Making (RIDM), the USNRC proposed the Intermediate Break LOCA (IBLOCA) scenario to become a design basis accident. These studies reported that the probability of a complete rupture of a pipe depends on the pipe size, and in particular that this probability is higher for smaller tubes. Therefore, the double guillotine break of smaller pipes connected to the primary side of a PWR such as the surge line or the ECCS lines should pose a more probable scenario than an integral rupture of the cold or hot legs. In order to assess the effectiveness of the safety strategies and the Emergency Operational Procedures (EOP) for a selected Nuclear Power Plant (NPP) and scenario, Best Estimates (BE) codes like RELAP5 are of great value because they allow simulating the overall behavior and response of the system under accident conditions. Furthermore, for licensing purposes, BE simulations should be combined with uncertainty analyses to yield the so-called Best Estimate Plus Uncertainty (BEPU) calculations. In the present paper a preliminary assessment of an IBLOCA BEPU calculation is carried out for the Asc ó -2 NPP (3-loop PWR Westinghouse design). The transient follows its reported EOPs with Emergency Core Cooling Systems (ECCS) failure assumptions for core uncover conditions. A preliminary evaluation of the effectiveness in the selection of the input uncertainty parameters is presented. Those that were concluded as relevant for LBLOCA and SBLOCA in previous works are analyzed to determine which are the most influential to be considered in IBLOCA. In addition, the influence of the nodalization qualification is also studied. Two different nodalizations are compared for assessing the significance of the modelling approach in BEPU analyses: the first one, a 1D vessel nodalization that was qualified with operational events reported in the actual NPP; and the second one, the same nodalization with a Pseudo 3D modelling of the vessel (core parallel channels with transversal flow paths and DC parallel channels with azimuthal connections). This second nodalization was qualified for SBLOCA accidents with experiments performed at Integral Test Facilities (G7.1 of PKL and Test 3 of LSTF ) b y the use scaling techniques (SCUP methodology).Postprint (published version

RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of test blanket modules involving helium flows into heavy liquid metal

The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. Thermal hydraulic safety analyses are being carried out with the system code RELAP/SCDAPSIM/MOD4.0 whic h was initially designed to predict the behaviour of light water reactor systems during normal and accident conditions. The code is being developed as part of the international SCDAP Development and Training Pr ogram (SDTP) coordinated by Innovative Systems Software (ISS). The modeling strategy of the RELAP code for the simulation of two-phase flows is based on a single-fluid two-phase approach with a set of momentum, energy and mass equations for each phase. The two phases are liquid-water and gas phase mixture of steam and non-condensable gases. Phase interactions, such as interphase friction and heat transfer, are modelled by closure relations based on experimental/numerical correlations that depend on the flow regime. In cooperation with ISS, the IPR team has implemented LLE liquid phase thermodynamic properties as a working fluid alternative to water and appropriate wall-to-LLE heat transfer correlations. However, in order to analyze some of the postulated off-normal events, there is a need to simulate the mixing of helium and Lead-Lithium fluids In the standard RELAP/SCDAPSIM/MOD4.0 ve rsion it is not possible to simulate a mixture of a non-water fluid with a non-condensable. In addition to that, t he available flow regime maps for vertical and horizontal flows in RELAP are specific for steam/water pair, which may not be suitable for LLE/helium pair. The Technical University of Catalonia is cooperating with IPR to adapt the RELAP/SCDAPSIM/MOD4.0 code to allow the si mulation of LLE and he mixture. This paper presents the results of the first step of the project, which includes a state of the art on simulation of liquid metals mixed with non-condensable using system codes, the implementation of the necessary code modifications to allow for a LLE/he mixture a nd preliminary results using the modified code version for horizontal and vertical configurations

Significance of the input parameters selection and the nodalization qualification in the final results of an IBLOCA BEPU calculation

In the framework of Design Basis Accidents (DBA) for Pressurized Water Reactor (PWR), and based on recent studies on pipe integrity combined with the Risk-Informed Decision Making (RIDM), the USNRC proposed the Intermediate Break LOCA (IBLOCA) scenario to become a design basis accident. These studies reported that the probability of a complete rupture of a pipe depends on the pipe size, and in particular that this probability is higher for smaller tubes. Therefore, the double guillotine break of smaller pipes connected to the primary side of a PWR such as the surge line or the ECCS lines should pose a more probable scenario than an integral rupture of the cold or hot legs. In order to assess the effectiveness of the safety strategies and the Emergency Operational Procedures (EOP) for a selected Nuclear Power Plant (NPP) and scenario, Best Estimates (BE) codes like RELAP5 are of great value because they allow simulating the overall behavior and response of the system under accident conditions. Furthermore, for licensing purposes, BE simulations should be combined with uncertainty analyses to yield the so-called Best Estimate Plus Uncertainty (BEPU) calculations. In the present paper a preliminary assessment of an IBLOCA BEPU calculation is carried out for the Asc ó -2 NPP (3-loop PWR Westinghouse design). The transient follows its reported EOPs with Emergency Core Cooling Systems (ECCS) failure assumptions for core uncover conditions. A preliminary evaluation of the effectiveness in the selection of the input uncertainty parameters is presented. Those that were concluded as relevant for LBLOCA and SBLOCA in previous works are analyzed to determine which are the most influential to be considered in IBLOCA. In addition, the influence of the nodalization qualification is also studied. Two different nodalizations are compared for assessing the significance of the modelling approach in BEPU analyses: the first one, a 1D vessel nodalization that was qualified with operational events reported in the actual NPP; and the second one, the same nodalization with a Pseudo 3D modelling of the vessel (core parallel channels with transversal flow paths and DC parallel channels with azimuthal connections). This second nodalization was qualified for SBLOCA accidents with experiments performed at Integral Test Facilities (G7.1 of PKL and Test 3 of LSTF ) b y the use scaling techniques (SCUP methodology)

Development of flow regime maps for lead lithium eutectic–helium flows

nstitute for Plasma Research (IPR), Gandhinagar (India) is currently involved in the design and development of its Lead-Lithium Ceramic Breeder (LLCB) module for testing in the International Thermo-nuclear Experimental Reactor (ITER). In order to fulfill the ITER safety requirements, some postulated events need to be analyzed. Among them, the internal loss of coolant accident is being studied using RELAP/SCDAPSIM/MOD4.0 code. To this aim, RELAP/SCDAPSIM/MOD4.0 capabilities of modeling lead lithium eutectic (LLE) and helium flows are being extended nowadays by a collaboration of three institutions, namely IPR, Innovative System Software (ISS) and UPC-BarcelonaTech. The current study is part of this effort and is focused on the adjustments of existing RELAP/SCAPSIM/MOD4.0 flow regime maps in order to deal with the LLE-helium pair. Due to the lack of experimental data, the current study is based on a numerical assessment of LLE-helium flows. The volume of fluid method implemented in OpenFOAM toolkit has been chosen to characterize the flow regime at different mixture velocities and helium fractions. The paper includes a description of the followed methodology and its validation. Results of the systematic simulations of both horizontal and vertical upward flows are shown and the proposed flow diagrams presented. It is worthwhile to mention that the numerically obtained flow regime maps must be validated by an experimental campaign. However, the current approach allows the use of RELAP/SCDAPSIM/MOD4.0 code to properly initiate the study of breeding blanket off-normal events postulated for Lead-Lithium Ceramic Breeder, Helium Cooled Lithium-Lead and Dual Coolant Lithium-Lead designs.Postprint (author's final draft