Giuseppe Vella

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

A new correlation for quench front velocity

Abstract The purpose of this paper is to present a new correlation for the prediction of the quench front velocity in the rewetting of hot dry surfaces by falling water films. This correlation is valid in the full range of the operating parameters where it proves very successful, providing a root mean square error of about 2.5%, with respect to accurate numerical solutions of the mathematical problem involved.

TRACE Code Analyses for the IAEA ICSP on “Integral PWR Design Natural Circulation Flow Stability and Thermo-Hydraulic Coupling of Containment and Primary System During Accidents”

Considering the world energy demand increase in order to fulfill an environmental and economic sustainability, the energy policy of each country has to diversify the sources of energy and use stable, safe energy production option able of producing electricity in a clean way contributing in cutting the CO2 emission. In the framework of the sustainable development, today the use of advanced nuclear power plant, have an important role in the environmental and economic sustainability of country energy strategy. In the last 20 years, in fact, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs considering also the use of natural circulation for the cooling of the core in normal and transient conditions. In this framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena characterizing the Multi-Application Small Light Water Reactor (MASLWR) design, a small modular integral pressurized light water reactor relying on natural circulation during both steady state and transient operation. It includes an integrated helical coil steam generator as well. Starting from an experimental campaign in support of the MASLWR concept design verification, the planned work, will be not only to specifically investigate the concept design further but also advance the broad understanding of integral natural circulation reactor plants and accompanying passive safety features as well. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System During Accidents” is hosting at OSU and the experimental data will be developed at the OSU-MASLWR facility. The purpose of this IAEA ICSP is to provide experimental data on single/two-phase flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type light water reactors. These data can be used to assess thermal hydraulic codes for reactor system design and analysis as well. The first planned test investigates a stepwise reduction in the primary mass inventory of the facility while operating at reduced power (decay power). The second planned test, investigates a loss of feed water transient with subsequent primary blowdown due to automatic depressurization system actuation and long term cooling phase. The target of this paper is to contribute to the thermal hydraulic analysis of the expected phenomena of these transients on the basis of the TRACE V5 Patch 01 calculated data developed during the double-blind phase of the ICSP.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

Analysis of Primary/Containment Coupling Phenomena Characterizing the MASLWR Design During a SBLOCA Scenario

Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011).