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.

The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

Supplemental Stress and Fracture Mechanics Analyses of Pressurized Water Reactor Pressure Vessel Nozzles

2012 ◽  
Author(s):  
Matthew Walter ◽  
Shengjun Yin ◽  
Gary L. Stevens ◽  
Daniel Sommerville ◽  
Nathan Palm ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Additional Stress ◽  
Corner Crack ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Reactor Pressure

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: • To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). • To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. • To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, “Fracture Toughness Requirements,” and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

Cracking in Alloy 600 Penetration Nozzles: A Regulatory Perspective

2004 ◽  
Author(s):  
Allen L. Hiser
Keyword(s):  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Alloy 600 ◽  
Pressurized Water ◽  
New Findings ◽  
Regulatory Perspective ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure ◽  
The U.S

Cracking in Alloy 600 penetration nozzles in the reactor pressure vessel upper heads of pressurized water reactors (PWRs) has been an issue in the U.S. nuclear industry for more than 10 years. This paper provides a regulatory perspective on how the U.S. industry and the U. S. Nuclear Regulatory Commission (NRC) have addressed new findings over the last 10 years and responded to ensure safety of U.S. PWRs. This paper provides a summary going back to the early 1990’s and emphasizes activities over the last three years since the discovery of circumferential cracking at Oconee, and provides a perspective regarding on-going NRC activities.

Impact of Using ASME Section XI Code Case N-830 on Plant Heatup and Cooldown Pressure-Temperature Limit Curves for Pressurized Water Reactors

2019 ◽  
Author(s):  
Alexandria Carolan ◽  
Benjamin Mays ◽  
Anees Udyawar ◽  
J. Brian Hall
Keyword(s):  
Lower Bound ◽  
Pressure Vessel ◽  
Master Curve ◽  
Temperature Limit ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors ◽  
The Impact ◽  
Reactor Pressure

Abstract Nuclear plant reactor pressure vessel heat-up and cool-down pressure temperature (P-T) limit curves are determined using ASME Section XI, Appendix G. ASME has adopted into ASME Section XI Appendix G the allowable use of the ASTM E1921 master curve fracture toughness based reference temperature (T0) to index the KIc curve. ASME Section XI Code Case N-830 allows the use of the KJc 95% lower bound master curve indexed using T0 directly. In ASME Section XI Appendix G, the equation RTT0 = T0 + 19.4°C, as an alternate RTNDT, shifts the KIc curve to approximate the KJc 95% lower bound master curve, however, the KIc exponential curve parameters are different. Thus, this paper evaluates the impact on plant heatup and cooldown pressure-temperature limit curves between the two ASME approved methods for typical pressurized water reactors (PWR). Different degrees of embrittlement are assessed to determine differences in the two approaches on reactor pressure vessel (RPV) beltline operating curves. Furthermore, in the proposed Revision 19 of Regulatory Guide 1.147, the US NRC has included a condition on the use of Code Case N-830 that prohibits the use of the current KIc equation in ASME Section XI Appendix G when these values are above the KJc lower bound 95% curve at temperatures below T0 − 64°C. This paper briefly discusses this NRC condition on the P-T limit curves.

Measurement Challenges Associated With Irradiated Reactor Components

1997 ◽  
Vol 503 ◽  
Author(s):  
Steven R. Doctor
Keyword(s):  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Reactor Pressure Vessel ◽  
Boiling Water Reactors ◽  
Background Information ◽  
Nondestructive Measurement ◽  
Pressurized Water ◽  
Radiation Fields ◽  
Water Reactors ◽  
Reactor Pressure

ABSTRACTNuclear reactor components are known to degrade as reactors age. The reactor pressure vessel and the reactor internals are subjected to the highest radiation fields and thus, age faster. Some boiling water reactors are experiencing extensive cracking of the core shroud and some pressurized water reactors are experiencing a reduction in their fracture toughness. This paper provides background information on reactor pressure vessel materials, fabrication, and construction practices. It identifies some of the challenges that are associated with the nondestructive measurement of changes in material properties associated with these components. The inspection environment, surface conditions and designs are also discussed.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

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