System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

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.

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PRESSURIZED WATER REACTOR-FULL LENGTH EMERGENCY COOLING HEAT TRANSFER (PWR- FLECHT) TESTS PROJECT.

10.2172/4148978 ◽

1970 ◽

Author(s):

J Haire ◽

G Brockett

Keyword(s):

Heat Transfer ◽

Full Length ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor

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Experiment and Numerical Simulation of Flow Mixing and Heat Transfer in Fuel Assembly for Pressurized Water Reactor

Transactions of the KSME C Industrial Technology and Innovation ◽

10.3795/ksme-c.2021.9.2.085 ◽

2021 ◽

Vol 9 (2) ◽

pp. 85-95

Author(s):

Wang Kee In

Keyword(s):

Heat Transfer ◽

Numerical Simulation ◽

Fuel Assembly ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Flow Mixing

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Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

Journal of Pressure Vessel Technology ◽

10.1115/1.4040790 ◽

2018 ◽

Vol 140 (5) ◽

Author(s):

Bipul Barua ◽

Subhasish Mohanty ◽

Joseph T. Listwan ◽

Saurindranath Majumdar ◽

Krishnamurti Natesan

Keyword(s):

Finite Element ◽

Stainless Steel ◽

Stress Analysis ◽

Nuclear Reactor ◽

Three Dimensional ◽

Fatigue Loading ◽

Cyclic Stress ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

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Estimation of Pressure Loads During Ex-Vessel Steam Explosion

Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance ◽

10.1115/icone16-48048 ◽

2008 ◽

Author(s):

Matjazˇ Leskovar

Keyword(s):

Heat Transfer ◽

Steam Explosion ◽

Phase 1 ◽

Interaction Process ◽

Radioactive Material ◽

Primary System ◽

Pressurized Water Reactor ◽

Reactor Accident ◽

Pressurized Water ◽

Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central and side melt pours), the cavity water sub-cooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplets freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

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