Davide Papini

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.

On Density Wave Instability Phenomena – Modelling and Experimental Investigation

Density Wave Oscillations (DWOs) are dealt with in this work as the most representative instabilities frequently encountered in the boiling systems. This dynamic type instability mode – resulting from multiple feedback effects between the flow rate, the vapour generation rate and the pressure drops in the boiling channel – constitutes an issue of special interest for the design of industrial systems and equipments involving vapour generation (Yadigaroglu, 1981). In the nuclear area, instability phenomena can be triggered both in Boiling Water Reactor (BWR) fuel channels (where they are moreover coupled through neutronic feedbacks with the neutron field), and in steam generators, which experience boiling phenomena inside parallel tubes. The latter is typical configuration of all the once-through steam generators, considered in this work with respect to integral Smallmedium Modular Reactors (SMRs)1 applications. Extensive attention is required because parallel channel instability is very difficult to be immediately detected when occurs in steam power systems, being the total mass flow of the system stable while the instability is locally triggered among some of the parallel channels. Thermally induced oscillations of the flow rate and system pressure are undesirable, as they can cause mechanical vibrations, problems of system control, and in extreme cases induce heat transfer surface burn-out. Large amplitude fluctuations in the heater wall temperature (so named thermal oscillations) usually occur under DWO conditions. Continual cycling of the wall temperature can lead to thermal fatigue problems which may cause tube failure. It is clear from these examples that the flow instabilities must be avoided in the design and operation of the various industrial systems. The safe operating regime of a two-phase heat exchanger can be determined by instability threshold values of system parameters such as flow rate, pressure, inlet temperature and exit quality. To the aim, both basic experiments and numerical analyses are necessary. This work is dedicated to the study (from theoretical, numerical and experimental point of view) of density wave phenomena, aimed at instability threshold prediction, DWO characterization and linear stability analysis as well.

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Decay Heat Removal (DHR) is a fundamental safety function which is often accomplished in the advanced LWRs relying on natural phenomena. A typical passive DHR system is the two-phase flow, natural circulation, closed loop system, where heat is removed by means of a steam generator or heat exchanger, a condenser, and a pool. Different condenser tube arrangements have been developed for applications to the next generation NPPs. The two most used configurations, namely, horizontal and vertical tube condensers, are thoroughly investigated in this paper. Several thermal-hydraulic features were explored, being the analysis mainly devoted to the description of the best-estimate correlations and models for heat transfer coefficient prediction. In spite of a more critical behaviour concerning thermal expansion issues, vertical tube condensers offer remarkably better thermal-hydraulic performances. An experimental validation of the vertical tube correlations is provided by PERSEO facility (SIET labs, Piacenza), showing a fairly good agreement.

Experimental Characterization of a Passive Emergency Heat Removal System for a GenIII

Among the several types of passive safety systems adopted in new generation reactor designs, the experimental investigation of a closed loop, two-phase flow, natural circulation system is depicted. Emergency Heat Removal Systems (EHRSs) based on this solution are envisaged as safety-engineered features for advanced nuclear reactors, as in the IRIS reactor. An experimental facility simulating one EHRS-like loop has been built and operated at SIET labs in Piacenza (Italy). The facility is a natural circulation, sliding pressure, and electrically heated loop, with a helical coil steam generator as a heat source and a horizontal tube pool condenser as a heat sink. A steady-state analysis is provided to characterize the system behaviour and its key parameters. Because of the loop limited volume, oscillations of the main parameters (temperatures, flowrate, pressure) may be expected. The oscillating phenomena detected during the experimental campaign are discussed; a reasonable explanation is at last proposed.

Evaluation of the PAR Mitigation System in Swiss PWR Containment Using the GOTHIC Code

Abstract The installation of passive autocatalytic recombiners (PARs) in the containment of operating nuclear power plants (NPPs) is increasingly based on three-dimensional studies of severe accidents that accurately predict the hydrogen pathways and local accumulation regions in containment and examine the mitigation effects of the PARs on the hydrogen risk. The GOTHIC (Generation Of Thermal-Hydraulic Information for Containments) code is applied in this paper to study the effectiveness of the PARs installed in the Gösgen NPP in Switzerland. A fast release of a mixture of hydrogen and steam from the hot leg during a total station blackout is chosen as the limiting scenario. The PAR modeling approach is qualified simulating two experiments performed in the frame of the OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) THAI (Thermal-hydraulics, Hydrogen, Aerosols and Iodine) project. The results of the plant analyses show that the recombiners cannot prevent the formation of a stratified cloud of hydrogen (10% molar concentration), but they can mitigate the hydrogen accumulation once formed. In the case of the analyzed fast release scenario, which is characterized by increasing loads with large initial flow rate and high hydrogen concentration values, it is shown that, when a large number of recombiners are installed, the global outcome in relation to the combustion risk does not depend on the details of the single PAR behavior. The hydrogen ignition risk can be fully mitigated in a timeframe ranging from 15 to 30 min after the fast release, according to the dependence of the PAR efficiency model on the adopted parameters.

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