Thermal-hydraulic code RELAP5, originally developed by INEEL under NRC contract for LWRs LOCA analysis, has been extensively validated and it is actually used worldwide as a best estimate code for LWRs transient analysis. This code was chosen in the early 2000s as the reference code for the thermal-hydraulics and accident analysis of Heavy Liquid Metal Fast Reactors within the frame of Italian research programs. Several original RELAP5 routines have been updated in order to implement suitable correlations generating the reference physical and thermodynamic properties for Lead and Lead-Bismuth Eutectic fluids. Similarly, some specific heat transfer correlations for liquid metals have been added. In almost twenty years, ENEA has conducted an extensive validation program to demonstrate the capability of the code to simulate the thermal-hydraulic behaviour of heavy liquid metal systems. This program was mainly conducted on the facilities of the Brasimone center (CHEOPE, CIRCE, NACIE), but also used data from international programs (MEGAPIE International Experiment and LACANES Benchmark OECD/ NEA). Furthermore, the participation of ENEA in the projects financed by the European Commission in several Euratom programs for the study and design of HLMcooled reactors (ELSY, SEARCH, MAXSIMA, LEADER, SESAME. MYRTE) has allowed to increase the confidence on the transient and accidental analyses conducted with RELAP5 also through the comparison with the results of other codes. RELAP5 is currently a reliable tool both as a support to reactor design (safety by design) and for accidental analyses.

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.

Passive systems are being implemented in several nuclear plants already constructed, under construction or under design. The purpose of these systems is to increase the inherent safety of the plant in any transient condition, so they can provide a higher safety level in case of an accident. Best estimate thermal-hydraulic system codes, widely adopted in the nuclear field to perform deterministic safety analyses, need to be validated against the phenomena typical of passive systems (e.g. natural circulation, thermal-hydraulic behavior of large pools, low pressure phenomena, etc.). An experimental test facility called PERSEO (in-Pool Energy Removal System for Emergency Operation) was built at SIET laboratory in Piacenza (Italy) modifying the existing PANTHERS IC-PCC facility. It is a full-scale test facility aimed at studying a new passive heat removal system and it can be used to analyze the capability of system codes to simulate the related phenomena. In the present analysis, Test 9, which was performed in the PERSEO facility at around 4.1 MPa to study the activation and the heat removal capability of the proposed system, has been simulated with the RELAP5-3D code to assess its capability to predict the dominant phenomena of the transient. Particular attention has been dedicated to the analysis of condensation heat transfer coefficient computed by the code, since in previous analyses it has been found an evident underestimation of the power exchanged by the passive system. Alternative correlations for the condensation heat transfer coefficient have been identified and a correction factor for the heat transfer coefficient calculated by RELAP5-3D has been computed. After the application of the correction factor, the set composed by Nusselt-Kutateladze-Chen correlations provided the better agreement with the experimental data. Finally, the accuracy of the code calculation has been evaluated qualitatively and quantitatively by adopting the Fast Fourier Transform Based Method.

Advanced small LWRs, are considered as one of the key design options for the development of a safer nuclear technology considering their inherent safety due to the adoption of passive mitigation strategies and lower nominal power. In the framework of deterministic safety analyses for Small Modular Reactors (SMRs), a model of a generic IRIS SMR was developed by using the severe accident code ASTEC (Study carried out with ASTEC V2, IRSN all rights reserved, [2019]). The ASTEC code thermal-hydraulics modules have been used for the reactor modelling, and the nodalization approach has been described in the present work. The objective of the paper is to analyze the code capability and the consequent applicability to model an integral-type reactor and to simulate the complex thermal-hydraulic phenomena occurring in a passive mitigation strategy. The analysis is based on 2-inches guillotine Direct Vessel Injection (DVI) line break transients. (C) 2021 Elsevier Ltd. All rights reserved.

The aim of this paper is to present a fractional scaling analysis (FSA) application for a system with interacting components where multiple figures of merits need to be respected during complex transient accident scenario with several consecutive time sequences. This paper presents FSA application to the International Reactor Innovative and Secure (IRIS) reactor and Simulatore Pressurizzato per Esperienze di Sicurezza 3 (SPES3) integral effects test (IET) facility. The FSA was applied for the small break loss of coolant accident (SBLOCA) on the direct vessel injection (DVI) line as the most challenging transient scenario. The FSA methodologies were applied for two figures of merits: (1) reactor and containment vessels pressure responses, and (2) reactor vessel water collapsed level response. The space decomposition was performed first. The reactor vessel and containment vessel were divided in components so that important phenomena and their consequences can be evaluated in each of them. After that, the time decomposition in consecutive time sequences was performed for the considered transient (DVI SBLOCA) based on the starts, or ends, of the defining events. The configuration of the system in each time sequence might be different and dependent on the control system actions connecting, or disconnecting, various components of the system due to the valves openings, or closings. This way, the important phenomena and their consequences can be evaluated for each component and time sequence. Also, this paper presents and discusses options for deriving nondimensional groups and calculation of distortions between prototype and model responses for complex transients containing multiple consecutive time sequences. The input data for scaling analysis are based on the results of RELAP/GOTHIC analysis performed for IRIS and RELAP analysis performed for SPES3. The scaling analysis was applied iteratively several times for different IRIS and SPES3 configurations. Based on the intermediate results, some components in the IRIS and SPES3 were redesigned so that the distortions between IRIS and SPES3 responses are decreased.

In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool.The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient.In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal.Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend of temperature stratification in the LBE pool during both forced and natural convection conditions is rather well reproduced by the codes. The largest uncertainties of the simulation concern the DHR heat exchanger behavior, which showed large instabilities in its operation during the tests. (C) 2014 Elsevier B.V. All rights reserved.

As one of the Generation-IV reactor concepts, lead-alloy-cooled advanced nuclear energy systems (LACANES) have been studied worldwide in order to utilize the advantages of good heat transfer properties, neutron transparency and chemical inertness with air and water. Since the Fukushima accident, the passive safety aspect of the LACANES is increasingly emphasized due to outstanding natural circulation capability. To investigate the thermal-hydraulic capability of LBE, an international cooperation has been performed under OECD/NEA program, under the guidance of the Nuclear Science Committee by a task force named as Lead Alloy Cooled Advanced Nuclear Energy Systems (LACANES) since 2007. This international collaboration had dealt with computational benchmarking of isothermal LBE forced convection tests in the phase I, and the working group published a guideline for using one-dimensional system codes to simulate LBE forced circulation test results from HELIOS loop. The phase II was started after that, to give an additional guideline in the case of natural circulation. NACIE, one of benchmarking targets for the phase II which is a rectangular-shape loop located at ENEA-Brasimone Research Centre, Italy. NACIE test results were benchmarked by each participant using their one-dimensional thermal-hydraulic codes, and they are to follow the guideline from the LACANES phase I for regions where hydraulic loss occurs. Due to the selection of hydraulic loss coefficient relations by users, the cross-comparison results of international participants showed some discrepancies and the estimated mass flow rates had 13% of maximum error. Also, the future R&D areas are identified.

An accurate three-dimensional simulation of all the components of the primary circuit of a LFR (Lead Fast Reactor) cannot be performed with the current computational power One strategy to deal with such complex systems is to adopt a multi-scale approach, where different models and geometric representations are introduced for different parts of the reactor. This paper presents a preliminary assessment of a methodology developed in the framework of the FEM-LCORE code to simulate an accident scenario where natural circulation plays a key role in the heat removal.

Il presente documento riporta le attivita svolte nell’ambito della Linea Progettuale 1 (LP1), obiettivo B (Metodologie avanzate per la valutazione delle conseguenze incidentali), task B.2, del PAR 2012, ADP ENEA-MSE. Lo studio presenta una metodologia del tipo “risk-informed” per la analisi di rischio di impianti nucleari a fronte di eventi esterni, cosi come emerge dall’incidente di Fukushima. L’approccio proposto integra gli aspetti probabilistici con quelli deterministici per perfezionare gli strumenti attualmente in uso ai fini dell’analisi di sicurezza. Un’ applicazione pilota, in termini di definizione di sequenza dell'incidente e relativa valutazione e indicativamente proposta come significativo caso di studio. In particolare, gli incidenti di perdita di alimentazione elettrica alla centrale (o “Stazione Black-Out”) causati da eventi esterni, come tifoni e tornado, vengono analizzati sia dal punto di vista deterministico che da quello probabilistico.

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants.The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA.The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents.Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal.The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes.The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.

An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

This paper is part of the validation process of the system code, CATHARE 2, applied to Gas Cooled Reactor studies. In the frame of the European Integrated Project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation) a benchmark exercise on the experimental transients conducted in the HE-FUS3 helium facility (Brasimone. Italy) was organized to assess the codes capability in simulating the thermal-hydraulic and dynamic behavior of Very High Temperature Reactors (VHTR).HE-FUS3 loop contains a wide range of components characteristic of a VHTR like compressor, pipes, diffusers, valves, heaters and heat exchangers. Even if the loop characteristics/configuration is not prototypical of a VHTR design, the loop is operating with helium, the design pressure and temperature set respectively at 10.5 MPa and 530 degrees C.Experimental data for several steady tests and for two transients corresponding to Loss of Flow Accidents were available.This paper includes first a brief description of the CATHARE 2 code applied to Gas Cooled Reactors. Then a description of the CATHARE 2 modeling for each of the relevant components of the HE-FUS3 loop is given. CATHARE 2 calculations results are then compared and analyzed on the basis of the experimental data. Sensitivity calculations to the modeling and to some input parameters are performed and also described in this report. Taking into account the noted experimental uncertainties of some circuit data, the CATHARE 2 results are in a satisfactory agreement with the HE-FUS3 experiment for all the steady state and the transient tests. Needs for future code developments and for the experimental loop improvements are suggested. (C) 2011 Elsevier B.V. All rights reserved.

As highly promising coolant for new generation nuclear reactors, liquid Lead–Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations – ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI – contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal–hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.