An Innovative Approach to Nuclear Reactor Design Certification

1988 ◽  
Vol 110 (2) ◽  
pp. 161-165
Author(s):  
N. W. Brown ◽  
G. L. Gyorey ◽  
B. K. Genetti ◽  
M. A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Power Reactor ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
General Electric ◽  
Test Program ◽  
New Approach ◽  
The Us ◽  
Regulatory Commission ◽  
Made In

General Electric has proposed that the US Nuclear Regulatory Commission (NRC) consider adding an Appendix to 10CFR50 that would specifically address NRC Safety Review and Design Certification of advanced reactors through use of an experience building test program. The proposal was made in conjunction with the Department of Energy (DOE)-sponsored review of the General Electric advanced Liquid Metal Reactor (LMR) concept, Power Reactor Inherently Safe Module (PRISM). This paper provides a description of the proposed new 10CFR50 Appendix. It also provides the basis for the proposed new approach to Design Certification and outlines the plans that are in place for further review and consideration by the NRC.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Codes, Standards and Regulatory Topics for Nuclear Facilities

2008 ◽  
Author(s):  
Taunia Wilde ◽  
Shannan Baker ◽  
Gary M. Sandquist
Keyword(s):  
Quality Control ◽  
Quality Assurance ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Quality Assurance Program ◽  
Nuclear Facilities ◽  
The Us ◽  
Regulatory Commission

The design, construction, operation, maintenance, and decommissioning and decontamination of nuclear infrastructure particularly nuclear power plants licensed in the US by the US Nuclear Regulatory Commission (NRC) or operated by the US Department of Energy (DOE) or the US Department of Defense (DOD) must be executed under a rigorous and documented quality assurance program that provides adequate quality control and oversight. Those codes, standards, and orders regulate, document and prescribe the essentials for quality assurance (QA) and quality control (QC) that frequently impact nuclear facilities operated in the US are reviewed and compared.

scholarly journals Review of Practice for Deeply Embedded/Buried NPP Structures Subject to Seismic Loadings

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
National Laboratory ◽  
Earth Pressure ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Nuclear Containment ◽  
The Us ◽  
Reactor Building ◽  
Seismic Loadings ◽  
Knowledge And Practice ◽  
Regulatory Commission

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.

Fracture Toughness Properties of SA540 Steels for Nuclear Bolting Applications

1977 ◽  
Vol 99 (3) ◽  
pp. 419-426
Author(s):  
R. R. Seeley ◽  
W. A. Van Der Sluys ◽  
A. L. Lowe
Keyword(s):  
Fracture Toughness ◽  
Impact Energy ◽  
Nuclear Reactor ◽  
Charpy Impact ◽  
Nuclear Regulatory Commission ◽  
Charpy Impact Energy ◽  
Energy Correlation ◽  
Impact Properties ◽  
Minimum Requirements ◽  
Regulatory Commission

Large bolts manufactured from SA540 Grades B23 and B24 are used on nuclear reactor vessels and require certain minimum mechanical properties. A minimum fracture toughness of 125 ksi in. (137 MPa m) at maximum operating stresses is required by the Nuclear Regulatory Commission for these bolts. This minimum toughness property was determined by a stress analysis of a bolt. Minimum required Charpy impact properties were calculated by a fracture toughness-Charpy impact energy correlation and the minimum calculated fracture toughness. The fracture toughness, yield strength and Charpy V notch impact properties were determined for five commercial heats of SA540 steels. Correlations between the fracture toughness and Charpy impact properties of these materials were evaluated. The toughness-impact energy correlation used to set the minimum required Charpy impact properties was found to be unduly conservative, and a different correlation of these properties is suggested. The SA540 steels investigated exhibited fracture toughness properties in excess of the NRC minimum requirements.

Licensing of the ES-3100 Type B Shipping Container: Replacement of the DOT 6M Container

2005 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Dean R. Tousley ◽  
James C. Anderson
Keyword(s):  
State Of The Art ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Shipping Container ◽  
Nuclear Security ◽  
Material Transportation ◽  
Enriched Uranium ◽  
Regulatory Commission ◽  
The U.S

The National Nuclear Security Administration (NNSA) is shipping bulk quantities of fissile materials for disposition purposes, primarily highly enriched uranium (HEU), over the next 15 to 20 years. The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material. However, the 6M does not conform to the safety requirements in the Code of Federal Regulations (10 CFR 71[1]) and, for that reason, is being phased out for use in the secure transportation system of the U.S. Department of Energy (DOE) in early 2006. BWXT Y-12 is currently developing the replacement for the DOT 6M container for NNSA and other users. The new package is based on state-of-the-art, proven, and patented technologies that have been successfully applied in the design of other packages. The new package will have a 50% greater capacity for HEU than the 6M, and it will be easier to use with a state-of-the-art closure system on the containment vessel. This new package is extremely important to the future of fissile, radioactive material transportation. An application for license was submitted to the U.S. Nuclear Regulatory Commission (NRC) in February 2005. This paper reviews the license submittal, the licensing process, and the proposed contents of this new state-of-the-art shipping container.

Justification for Elevated Hydrogen Limits in Risk-Based Radioactive Waste Packages

2005 ◽  
Author(s):  
Russell Wagner
Keyword(s):  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Transportation Safety ◽  
Elevated Concentration ◽  
Type B ◽  
Hanford Site ◽  
Regulatory Commission ◽  
Concentration Limits ◽  
The U.S

The U.S. Nuclear Regulatory Commission (NRC) has provided set guidance that hydrogen concentrations in radioactive material packages be limited to 5 vol% unless the package is designed to withstand a bounding hydrogen deflagration or detonation. The NRC guidance further specifies that the expected shipping time for a package be limited to one-half the time to reach 5 vol% hydrogen. This guidance has presented logistical problems for transport of retrieved legacy waste packages on the Department of Energy (DOE) Hanford Site that frequently contain greater than 5 vol% hydrogen due to their age and the lack of venting requirements at the time they were generated. Such packages do not meet the performance-based criteria for Type B packaging, and are considered risk-based packages. Duratek Technical Services (Duratek) has researched the true risk of hydrogen deflagration and detonation with closed packages, and has developed technical justification for elevated concentration limits of up to 15 vol% hydrogen in risk-based packages when transport is limited to the confines of the Hanford Site. Duratek has presented elevated hydrogen limit justification to the DOE Richland Operations Office and is awaiting approval for incorporation into the Hanford Site Transportation Safety Document. This paper details the technical justification methodology for the elevated hydrogen limits.

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