Recent Revisions of SRP 3.7 and 3.8

2015 ◽  
Author(s):  
Ikbal Lebbi ◽  
Javad Moslemian
Keyword(s):  
Structural Design ◽  
Review Process ◽  
Seismic Analysis ◽  
Nuclear Regulatory Commission ◽  
Technical Issues ◽  
Technical Challenges ◽  
Regulatory Commission ◽  
Technical Guidance

During U.S. Nuclear Regulatory Commission (NRC) reviews of Design Certifications (DC) and Combined License (COL) applications, the NRC staff identified several technical issues related to seismic analysis and structural design of the containment and other seismic Category I structures and foundations. These technical issues resulted in a need of improved technical guidance to facilitate reviews of future DC and COL applications. NUREG-0800, “Standard Review Plan (SRP)”, Sections 3.7 and 3.8 provide guidance related to the review of seismic analysis and structural design. NUREG-0800 SRP Sections 3.7 and 3.8 have been revised recently in an effort to enhance NRC staff guidance for the review of future DC and COL applications. SRP Sections 3.7 and 3.8 have not been revised since 2007. Recent revisions of these Sections have been completed in 2013, except for SRP 3.7.1, only a “DRAFT” revision was completed in 2013. The objective of this paper is to briefly describe the important technical issues that presented technical challenges to the applicants during the NRC staff review process of DC and COL applications, and the proposed enhancements to the specific Sections of SRP 3.7 and 3.8. The paper presents complete details of the recent revisions of NUREG-0800 SRP Sections 3.7.1, 3.7.2, 3.7.3, 3.8.1, 3.8.3, 3.8.4, and 3.8.5.

Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor Reviews

2011 ◽  
Author(s):  
Stewart L. Magruder
Keyword(s):  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water Reactor ◽  
Design Features ◽  
Pressurized Water ◽  
Analysis And Evaluation ◽  
Small Modular Reactor ◽  
Efficiency And Effectiveness ◽  
Regulatory Commission ◽  
Regulatory Controls

The U.S. Nuclear Regulatory Commission staff plans to apply a more integrated, graded approach to the review of small modular reactor (SMR) pre-application activities and design applications. The concept is to improve the efficiency and effectiveness of the reviews by focusing on safety significant structures, systems, and components (SSCs). The unique design features associated with SMRs and knowledge gained reviewing other passive reactor designs present opportunities to risk-inform the SMR design certification process to a greater extent than previously employed. The review process can be modified for SMR applications by considering the aggregate of regulatory controls pertaining to SSCs as part of the review and determining those regulatory controls which may supplement or replace, as appropriate, part of the technical or engineering analysis and evaluation. Risk insights acquired from staff reviews of passive LWR designs (i.e. AP1000, ESBWR) can also be incorporated into the review process. Further, risk insights associated with integral pressurized water reactor (iPWR) design features (i.e. Underground facilities impact on turbine missiles review) can be incorporated into the review process.

Overview of Current Practice for Analysis and Design of Independent Spent Fuel Storage Installation Pads

2019 ◽  
Author(s):  
Gunup Kwon ◽  
Khaled Ata
Keyword(s):  
Nuclear Power ◽  
Seismic Analysis ◽  
Storage System ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Loading Conditions ◽  
Spent Fuel ◽  
Analysis And Design ◽  
License Application ◽  
Regulatory Commission

Abstract Nuclear power plant spent fuels are initially stored in the spent fuel pool. Then, the water cooled fuels are transferred in a concrete or steel cask and transported outside of the Fuel Handling Building (FHB) or the Reactor Building (RB) for long term on site storage. The spent fuel casks are typically stored on a slab-on-grade pad. The slab-on-grade pad is designed according to the U.S. Nuclear Regulatory Commission NUREG-1536 and NUREG-1567. The two Standard Review Plans provide guidance to the regulators for the review of cask storage system license application. The ISFSI pad analysis and design have to consider various loading conditions, such as earthquake and tornado loadings as well as normal operating loading conditions. Seismic analysis of the ISFSI pad requires considering interaction between the pad and the supporting soil. Various cask loading configurations on the pad also have to be considered. Due to the lack of specific guidelines, many ISFSI pad designs show overly conservative reinforcement. This study provides guidelines and procedure for the design of the ISFSI pad that are typically used in the nuclear industry. It is considered that the guidelines and practices described in this study help design engineers understand general guidance provided in the NRC Standard Review Plans.

The Joint Commission (JCAHO) and Nuclear Pharmacy

1989 ◽  
Vol 2 (5) ◽  
pp. 302-305
Author(s):  
Henry M. Chilton
Keyword(s):  
Quality Assurance ◽  
Review Process ◽  
Pharmacy Practice ◽  
Nuclear Regulatory Commission ◽  
Joint Commission ◽  
Regulatory Agencies ◽  
Healthcare Organizations ◽  
Record Keeping ◽  
Regulatory Commission ◽  
The Joint Commission

One of the central themes of the Joint Commission on Accreditation of Healthcare Organizations (JCAHO) review process involves emphasis on quality assurance of provided services with the goal of high quality patient outcomes. Within nuclear pharmacy practice, quality assurance activities are planned, systematic actions involving radiopharmaceuticals and equipment (and associated record-keeping functions for each) as well as cognitive clinical services. Concerns for appropriate quality assurance activities are voiced by other regulatory agencies (such as the Nuclear Regulatory Commission) and echoed by the JCAHO. The present discussion will highlight selected areas of nuclear pharmacy practice as it pertains to quality assurance of radiopharmaceutical receipt, preparation, and distribution, as well as equipment usage associated with these functions.

A Review Process and a Database for Waste-Package Documents

1990 ◽  
Vol 212 ◽  
Author(s):  
Charles G. Interrante ◽  
Carla A. Messina ◽  
Anna C. Fraker
Keyword(s):  
Information Needs ◽  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Waste Package ◽  
Long Term Storage ◽  
High Level Nuclear Waste ◽  
Regulatory Commission ◽  
High Level ◽  
Term Storage

ABSTRACTThe work reported here is part of a program conducted by the Nuclear Regulatory Commission on the efficacy of proposed plans for radionuclide containment for long-term storage of high-level nuclear waste (HLW). An important element of that program is the review and evaluation of available literature on components of a waste package. A review process and a database have been developed and tailored to provide information quickly to an individual who has a question about a particular material or component of a waste package. The database is uniquely suited to serve as a guide to indicate special areas where data and information needs exist on questions related to radionuclide containment. Additions to the database are made as information becomes available, and this source is as current as the published literature. A description of the review process and the database is given.

High-Level Waste Package Licensing Considerations for Extrapolating test Data

1986 ◽  
Vol 84 ◽  
Author(s):  
T.C. Johnson ◽  
K.C. Chang ◽  
T.L. Jungling ◽  
L.S. Person ◽  
C.H. Peterson ◽  
...  
Keyword(s):  
Test Data ◽  
Nuclear Regulatory Commission ◽  
High Level Waste ◽  
Geologic Repository ◽  
Waste Package ◽  
Performance Predictions ◽  
Technical Issues ◽  
Regulatory Commission ◽  
High Level ◽  
General Standard

AbstractPrograms intended to provide supporting information for the high-level radioactive waste (HLW) repository program must consider the licensing requirements and the technical issues involved with extrapolation of short-term test data to periods of up to 10,000 years. The licensing requirements of the Nuclear Regulatory Commission (NRC), and the issues the NRC staff considers important for the development of predictive methods, are described. Because performance predictions of the geologic repository and particular components of the waste package must largely be based upon inference, a reasonable assurance, on the basis of the record before the Commission, is the general standard that will be required.

scholarly journals Structural Design Challenges in Design Certification Applications for New Reactors

2011 ◽  
Author(s):  
Manuel Miranda ◽  
Joseph Braverman ◽  
Xing Wei ◽  
Charles Hofmayer ◽  
Jim Xu
Keyword(s):  
Structural Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Design Parameters ◽  
Specific Design ◽  
Safety Issues ◽  
Standard Design ◽  
Regulatory Commission ◽  
Design Challenges

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early resolution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific design parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design challenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

Historical Perspectives and Insights on ACRS Review of AP1000 Design Certification

2013 ◽  
Author(s):  
Hossein Nourbakhsh ◽  
Weidong Wang ◽  
Harold Ray ◽  
Thomas Kress
Keyword(s):  
Advisory Committee ◽  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Historical Perspectives ◽  
Standard Design ◽  
Regulatory Commission ◽  
The U.S ◽  
Made In

The U.S. Nuclear Regulatory Commission (NRC) requires that each application for a standard design certification be referred to the Advisory Committee on Reactor Safeguards (ACRS) for a review and report on those portions of the application which concern safety. This paper begins with perspectives on the role of the ACRS in the design certification review process. It then summarizes the ACRS observations and recommendations made in the Committee’s reports during the AP1000 design certification reviews to date.

scholarly journals Seismic Analysis Issues in Design Certification Applications for New Reactors

2011 ◽  
Author(s):  
Richard Morante ◽  
Manuel Miranda ◽  
Jim Xu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Regulatory Commission ◽  
Design Parameters ◽  
Specific Design ◽  
Site Specific ◽  
Safety Issues ◽  
Standard Design ◽  
Regulatory Commission

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early resolution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific design parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

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