Xu Cheng

Yanhua Yang

Walter Ambrosini

Dino A. Araneo

Passive Safety Systems in Advanced PWRs

The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS) methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks&#39; formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

Franco Pierro

Dino Araneo

Giorgio Galassi

Francesco D&#39;Auria

Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

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&#38;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.

Mario Carelli

Lawrence Conway

Milorad Dzodzo

Andrea Maioli

Luca Oriani

Gary Storrick

Bojan Petrovic

Andrea Achilli

Gustavo Cattadori

Cinzia Congiu

Roberta Ferri

Marco Ricotti

Davide Papini

Fosco Bianchi

Paride Meloni

Stefano Monti

Fabio Berra

Davor Grgic

Graydon Yoder

Alessandro Alemberti

The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

Horizontal buoyant jet is a fundamental flow regime for hydrogen safety analysis in power industry. The purpose of this study is to develop a fast non-Boussinesq engineering model the horizontal buoyant round jets. Verification of this integral model is established with available experimental data and comparisons over a large range of density variations with the CFD codes GASFLOW. The model has proved to be an efficient engineering tool for predicting horizontal strongly buoyant round jets.

J. Xiao

J.R. Travis

W. Breitung

Non-Boussinesq Integral Model for Horizontal Turbulent Buoyant Round Jets

The CFD code GASFLOW solves the time-dependent compressible Navier-Stokes Equations with multiple gas species. GASFLOW was developed for nonnuclear and nuclear applications. The major nuclear applications of GASFLOW are 3D analyses of steam/hydrogen distributions in complex PWR containment buildings to simulate scenarios of beyond design basis accidents. Validation of GASFLOW has been a continuously ongoing process together with the development of this code. This contribution reports the results from the open posttest GASFLOW calculations that have been performed for new experiments from the OECD SETH Benchmark. Discussed are the steam distribution tests 9 and 9bis, 21 and 21bis involving comparable sequences with and without steam condensation and the last SETH test 25 with steam/helium release and condensation. The latter one involves lighter gas mixture sources like they can result in real accidents. The helium is taken as simulant for hydrogen.

Peter Royl

John R. Travis

Wolfgang Breitung

Jongtae Kim

Sang Baik Kim

GASFLOW Validation with Panda Tests from the OECD SETH Benchmark Covering Steam/Air and Steam/Helium/Air Mixtures

Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear
power development in China requires reactor concepts with high safety level. For the
near-term and medium-term, the Chinese government decided for advanced
pressurized water reactors with an extensive usage of passive safety systems. This
paper describes some important criteria and the development program of the Chinese
large-scale pressurized water reactors. An overview on representative research
activities and results achieved so far on passive safety systems in various institutions
is presented.

X. Cheng

Y. H. Yang

Y. Ouyang

H. X. Miao

Role of Passive Safety Systems in Chinese Nuclear Power Development

A three dimensional computation fluid dynamics (CFD) code, GASFLOW, is applied to analyze the hydrogen risk for Qinshan-II 
                  nuclear power plant (NPP). In this paper, the effect of spray modes on hydrogen risk in the containment during a large 
                  break loss of coolant accident (LBLOCA) is analyzed by selecting three different spray strategies, that is, without spray, with direct spray and with both direct and recirculation spray. A strong effect of spray modes on hydrogen distribution is observed. However, the efficiency of the 
                  passive auto-catalytic recombiners (PAR) is not substantially affected by spray modes. The hydrogen risk is significantly increased by the direct spray, while the recirculation spray has minor effect on it. In order to simulate more precisely the processes involved in the PAR operation, a new PAR model is developed using CFD approach. The validation shows that the results obtained by the model agree well with the experimental results.

Jinbiao Xiong

CFD Application to Hydrogen Risk Analysis  and PAR Qualification

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

Junli Gou

Suizheng Qiu

Guanghui Su

Douna Jia 

Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

The paper summarizes the lesson learned from theoretical and experimental activities performed at the University of Pisa, Pisa, Italy, in past decades in order to develop a general methodology of analysis of heat and mass transfer phenomena of interest for nuclear reactor applications. An overview of previously published results is proposed, highlighting the rationale at the basis of the performed work and its relevant conclusions. Experimental data from different sources provided information for model development and assessment. They include condensation experiments performed at SIET (Piacenza, Italy) on the PANTHERS prototypical PCCS module, falling film evaporation tests for simulating AP600-like outer shell spraying conditions, performed at the University of Pisa, experimental data concerning condensation on finned tubes, collected by CISE (Piacenza, Italy) in the frame of the INCON EU Project, and experimental tests performed in the CONAN experimental facility installed at the University of Pisa. The experience gained in these activities is critically reviewed and discussed to highlight the relevant obtained conclusions and the perspectives for future work.

W. Ambrosini

M. Bucci

N. Forgione

A. Manfredini

F. Oriolo

Experiments and Modelling Techniques for  Heat and Mass Transfer in Light Water Reactors

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations

Passive Safety Systems in Advanced PWRs

Passive Safety Systems in Advanced PWRs

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