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  1. 1
    Technical specifications for the Oak Ridge Research Reactor

    Technical specifications for the Oak Ridge Research Reactor

    Published 1986
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  2. 2
    ORECA-1 a digital computer code for simulating the dynamics of HTGR cores for emergency cooling analyses.

    ORECA-1 a digital computer code for simulating the dynamics of HTGR cores for emergency cooling analyses.

    Published 1976
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  3. 3
    General Electric Company analytical model for loss-of-coolant analysis in accordance with 10CFR50 appendix K. Volume II

    General Electric Company analytical model for loss-of-coolant analysis in accordance with 10CFR50 appendix K. Volume II

    Published 1976
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  4. 4
    Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Published 2015
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  5. 5
    Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Published 2015
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  6. 6
    TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    Published 2000
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  7. 7
    Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

    Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

    Published 2001
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  8. 8
    Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer block...

    Published 1996
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  9. 9
    Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

    Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

    Published 1995
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  10. 10
    Results of Semiscale Mod-2C small-break loss-of-coolant accident without HPI (high-pressure injection) (S-NH) experiment series

    Results of Semiscale Mod-2C small-break loss-of-coolant accident without HPI (high-pressure injection) (S-NH) experiment series

    Published 1987
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  11. 11
    Code option guideline improvement using comparisons of RELAP4/MOD6 with forced and gravity-feed reflood data. [PWR]

    Code option guideline improvement using comparisons of RELAP4/MOD6 with forced and gravity-feed reflood data. [PWR]

    Published 1978
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  12. 12
    Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina nuclear power plant.

    Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina nuclear power plant.

    Published 2002
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  13. 13
    Effects of dissolved gas and downstream geometry during blowdown of a subcooled liquid. [PWR]

    Effects of dissolved gas and downstream geometry during blowdown of a subcooled liquid. [PWR]

    Published 1978
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  14. 14
    Experimental verification of thermal shock analysis. [PWR]

    Experimental verification of thermal shock analysis. [PWR]

    Published 1977
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  15. 15
    EMERGENCY CORE COOLING. Report of Advisory Task Force on Power Reactor Emergency Cooling.

    EMERGENCY CORE COOLING. Report of Advisory Task Force on Power Reactor Emergency Cooling.

    Published 1968
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  16. 16
    Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis.

    Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis.

    Published 1977
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  17. 17
    Effects of corrosion upon the adequacy of the 105-KE and 105-KW emergency coolant backup system

    Effects of corrosion upon the adequacy of the 105-KE and 105-KW emergency coolant backup system

    Published 1964
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  18. 18
    High flux beam reactor technical specifications

    High flux beam reactor technical specifications

    Published 1973
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  19. 19
    Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants (Cooperative Agreement DE-FC03-99SF21902, Am. M004) Final Technical Report

    Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants (Cooperative Agreement DE-FC03-99SF21902, Am. M004) Final Technical Report

    Published 2003
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  20. 20
    The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel

    The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel

    Published 2000
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