Giacomino Bandini

Core Loss during a Severe Accident (COLOSS).

KFKI Atomic Energy Research Institute (AEKI), Hungary Electricité de France (EDF), France Ente per le Nuove Tecnologie, l’Energia e l’Ambiente (ENEA) Italy Framatome ANP, France Forschungszentrum Karlsruhe GmbH (FZK), Germany European Commission – JRC/IE, International European Commission – JRC/ITU, International Paul Scherrer Institut (PSI), Switzerland Framatome ANP Gmbh, Germany SKODA-UJP Praha a.s., Czech Republic Universidad Politécnica de Madrid (UPM), Spain Ruhr-Universität Bochum (RUB), Germany Universität Stuttgart (IKE), Germany University Lappeenranta, Finland

Investigation of Core Degradation (COBE).

Abstract The COBE project started in February 1996 and finished at the end of January 1999. The main objective was to improve understanding of core degradation behaviour during severe accidents through the development of computer codes, the carrying out of experiments and the assessment of the computer codes’ ability to reproduce experimental behaviour. A major effort was devoted to quenching behaviour and a substantial achievement of the project was the design and commissioning of a new facility for the simulation of quenching of intact fuel rods. Two tests, carefully scaled to represent realistic reactor conditions, were carried out in this facility and the hydrogen generated during the quenching process was measured using two independent measuring systems. The codes were able to reproduce the results in the first test, where little hydrogen was generated but not the second test, where the extra steam produced during quenching caused an invigorated Zircaloy oxidation and a substantial hydrogen generation. A number of smaller parametric experiments allowed detailed models to be developed for the absorption of hydrogen and the cracking of cladding during quenching. COBE also investigated other areas concerned with late-phase phenomena. There was no experimental activity – the work included code development and the analysis of experimental data available to the project partners – either from open literature or from other projects such as Phebus-FP. Substantial improvement was made in the codes’ ability to simulate heat transfer in debris beds and molten pools and increased understanding was reached of control rod material interactions, the swelling of irradiated fuel and the movement of molten material to the lower head.

Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

The development of a conceptual design of an industrial-scale transmutation facility (EFIT) of several 100 MW thermal power based on accelerator-driven system (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

Fuel Dispersion and Flow Blockage Analyses for the Myrrha-Fastef Reactor by Simmer Code

This work illustrates the 3D set up model and the results concerning the recent analysis of fuel dispersion in the MYRRHA-FASTEF reactor performed with SIMMER code within the EU-FP7 SEARCH Project. Under severe accidental conditions, the release of fuel in the primary system can occur in case of fuel rod clad failure and degradation. Two cases were therefore taken into account, an imposed fuel release to study key parameters which influence the dispersion phenomenon and a coolant flow blockage in a fuel assembly.The reactor was simulated by a 3D Cartesian geometry with 65×63×42 cell mesh. Steady-state and transient analyses were performed by SIMMER-IV. Steady-state analysis was performed in order to assess the correct operability of the code and of the model. The results were compared with the design values. The most significant results obtained for temperature trends and profiles, velocity and mass flow rate trends are reported. Transient results were also analysed, i.e. fuel dispersion transients were simulated, comparing the effect of fuel porosity on the fuel dispersion inside the pool. In addition, the effects of the release position and the fuel particle dimension on the dispersion phenomenon were also investigated.The final section of the paper describes the effects of a flow blockage on the core degradation and dispersion of fuel particles in the primary circuit of the MYRRHA reactor. This simulation, with fuel porosity equal to 5%, started after a preliminary steady state condition. The mass flow rate in one of the inner fuel assemblies was then reduced to about 10% of the initial value.The results show that the SIMMER-IV code is capable of predicting steady-state results in good agreement with the nominal values, also confirming the correctness of the set up model.Copyright

ICONE23-2041 FIRST ANALYSIS OF AGS0, LT2 AND E9 CABRI TESTS WITH THE NEW SFR SAFETY CODE ASTEC-NA

23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

ICONE23-1980 POST-TEST OF THE SGTR EXPERIMENTAL TESTS CARRIED OUT IN THE LIFUS FACILITY

23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

Post Test Analysis of ICE Tests

Since 1999 ENEA is developing the heavy liquid metal (HLM) technology aiming to support the design and implementation of a Lead cooled Fast Reactor (LFR) and an Accelerator Driven System (ADS), both in the frame of the Italian and European research programs.In these contexts several experiments have been performed, in different fields, going from coolant thermal-hydraulic, component development and structural material characterization.Recently, in the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of a pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE).This experimental activity, named “Integral Circulation Experiment (ICE)”, has been implemented thanks a joint effort of several research institutes, mainly ENEA and University of Pisa, allowing to design an appropriate test section. This has been installed in the CIRCE facility, the largest worldwide experimental facility for the HLM technology investigation.The goal of the experiments was to demonstrate the technological feasibility of a heavy liquid metal (HLM) pooltype nuclear system in a relevant scale (1 MW), investigating the related thermal–hydraulic behavior under both steady state and transient conditions.This paper reports a description of the experiment, as well as the results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test, an un-protected loss of heat sink (ULOH) test, and an un-protected loss of flow (ULOF) test.The post-test analyses of the experiments is presented. The whole domain has been modeled by a suitable 1-D nodalization, and the results carried out have been studied performing numerical calculations by the REALP5 system code modified to take in account the LBE thermal-physical properties when employed as nuclear coolant.The obtained experimental results as well as the performed post-test analysis have demonstrated the thermal-hydraulic and technological feasibility of a pool-type nuclear system cooled by HLM.Copyright

Validation of CATHARE V2.5 Thermal-Hydraulic Code Against Full-Scale PERSEO Tests for Decay Heat Removal in LWRs

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.Copyright

Analysis of Protected Accidental Transients in the EFIT Reactor With the RELAP5 Thermal-Hydraulic Code

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of the European EUROTRANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside the reactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor in safe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system.Copyright