Paride Meloni

The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

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.

SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

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.

Assessment of RELAP5 MOD3.3 and CATHARE2 V1.5A against a Full Scale Test of PERSEO Device

PERSEO device was developed 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 device was tested at SIET Thermal-hydraulic Research Centre by modifying the existing PANTHERS IC-PCC facility. Two types of tests were performed: integral tests and stability tests. The experimental data acquired in the test campaign allowed a validation of a RELAP5/mod 3.3 beta release and CATHARE2 V1.5a/Mod8.1 full scale model of the PERSEO device. The paper deals with the comparison between the two codes against an integral test considered representative from the point of view of the PERSEO functioning and it highlights capabilities and limits of the codes in simulating such kind of test.Copyright

Cross-Comparison of One-Dimensional Thermal-Hydraulic Codes on Natural Circulation Analysis of NACIE Loop Test for Lead-Alloy Cooled Advanced Nuclear Energy Systems (LACANES)

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.Copyright

Preliminary Results on the Coupling of a Three-Dimensional Lead Fast Reactor Model and a One-Dimensional External Loop

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.Copyright

TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

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.© 2012 ASME

The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

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.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