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Comprehensive Analyses of Nuclear Safety System Codes

[+] Author Affiliations
Glenn A. Roth, Fatih Aydogan

University of Idaho, Idaho Falls, ID

Paper No. IMECE2013-63773, pp. V06BT07A048; 29 pages
doi:10.1115/IMECE2013-63773
From:
  • ASME 2013 International Mechanical Engineering Congress and Exposition
  • Volume 6B: Energy
  • San Diego, California, USA, November 15–21, 2013
  • Conference Sponsors: ASME
  • ISBN: 978-0-7918-5629-1
  • Copyright © 2013 by ASME

abstract

Many nuclear system codes have been developed for the main purpose of analyzing reactor performance of a nuclear power plant system during steady state and transient conditions. These codes generally include power plant component models for pumps, pipes, steam generators, pressurizers and other components. The parallel development of these nuclear system codes has been supported by government laboratories, universities, private entities and other organizations throughout the world. This has resulted not only in multiple codes, but multiple versions of the same code with different capabilities. The development paths of each code version have been driven by specific needs. The challenge for the user is to select a code that performs well for the desired analysis problem. Therefore, this work compares different aspects of various nuclear system codes. Firstly, it compares the governing equations for mass, momentum and energy in the evaluated system codes. Secondly, it compares all the codes’ closure models. Closure models are used in system codes to model thermal and mechanical non-equilibrium as well as the coupling of the phases. Thirdly, it compares the Separate Effect Tests (SET) and Integral Effect Tests (IET) employed for the verification and validation (V&V) during the development of each system code. These comparisons cover several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. Fourthly, major assumptions about the governing and closure equations in these codes are compared and discussed. Fifthly, numerical approach of every code is benchmarked with each other since numerical approach not only affects the speed of the system codes but also the accuracy of the results. Sixthly, the limitations of the codes are evaluated because these codes are challenged by analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, and many other design features that may not be fully analyzed by current system codes. The results of this work serve as a guide for development of these system codes and indicate areas where models must be improved to adequately address issues with new reactor design and development activities.

Copyright © 2013 by ASME

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