scholarly journals Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

1998 ◽  
Author(s):  
J.A. Wang ◽  
F.B.K. Kam
Keyword(s):  
Data Base ◽  
International Atomic Energy Agency ◽  
Pressure Vessel ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
International Database ◽  
Energy Agency ◽  
Oak Ridge ◽  
Regulatory Commission ◽  
Reactor Pressure

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.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

2006 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Reactor Pressure Vessel ◽  
High Dose ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Oak Ridge ◽  
Reactor Pressure

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

Fracture Toughness of Stainless Steel Cladding for Evaluation of the Degraded Davis-Besse RPV Head

2006 ◽  
Author(s):  
Randy K. Nanstad ◽  
Mikhail A. Sokolov
Keyword(s):  
Fracture Toughness ◽  
Stainless Steel ◽  
Pressure Vessel ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Shear Mode ◽  
Fracture Toughness Test ◽  
Test Results ◽  
Acid Attack ◽  
Oak Ridge

Boric acid attack in the reactor pressure vessel (RPV) head of the Davis-Besse (D-B) nuclear plant led to wastage through the 150-mm low alloy steel head such that the stainless steel cladding was exposed. The Heavy-Section Steel Technology (HSST) Program at Oak Ridge National Laboratory was commissioned by the Nuclear Regulatory Commission to conduct a program of testing and analysis to enable an evaluation of the structural significance of cladding defects found in the wastage cavity of the D-B head. The overall test program consisted of material characterization at 316°C (600°F) of cladding materials, pressure vessel burst tests of cladding discs with and without flaws, and extensive analytical studies. Three different cladding materials were tested and evaluated, one from an unused commercial RPV that was used for the clad-burst experiments, an archival cladding previously used for various experimental and irradiation experiments, and the cladding from the D-B head. This paper compares and discusses the fracture toughness test results conducted with the three claddings, and the fractographic analyses conducted on the clad-burst discs. Comparison of J-resistance curves for the three clad materials shows significant material variability and disparity in the results from two test specimen types. Fractographic examinations of clad-burst discs showed transition from ductile tearing to shear mode of fracture. The relationship of the cladding test results with the clad-burst results is discussed.

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

2015 ◽  
Author(s):  
Gary L. Stevens ◽  
Mark T. Kirk ◽  
Terry Dickson
Keyword(s):  
Fracture Mechanics ◽  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Stress Concentrations ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Regulatory Issue ◽  
Reactor Pressure

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

A Risk-Informed Assessment of ASME Section XI, Appendix E

2011 ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Evaluation Criteria ◽  
Nuclear Regulatory Commission ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
American Society ◽  
Regulatory Commission ◽  
Reactor Pressure ◽  
The U.S

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Non mandatory Appendix E, “Evaluation of Unanticipated Operating Events”, provides a deterministic procedure for evaluating reactor pressure vessel (RPV) integrity following an unanticipated event that exceeds the operational limits defined in plant operating procedures. The recently developed risk-informed procedure for Appendix G to Section XI of the ASME Code [2, 3], and the development by the U.S. Nuclear Regulatory Commission (NRC) of the alternate Pressurized Thermal Shock (PTS) rule [4, 5, 6] led to initiation of this study to determine if the Appendix E evaluation criteria are consistent with risk-informed acceptance criteria. The results of the work presented in this paper demonstrate that Appendix E is consistent with risk-informed criteria developed for PTS and Appendix G and ensures that evaluation of RPV integrity following an unanticipated event would not violate material or operational limits recently defined using risk-informed criteria. Currently, Appendix E does not have evaluation criteria for BWR vessels; however, as part of this study, risk-informed analyses were performed for unanticipated heat-up events and isothermal, overpressure events in BWR plant designs.

Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1

2010 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Computational Study ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Problem Class ◽  
Wide Range ◽  
Significant Generalization ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.

Distributions of Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluations

2002 ◽  
Author(s):  
F. A. Simonen ◽  
G. J. Schuster ◽  
S. R. Doctor ◽  
T. L. Dickson
Keyword(s):  
Pacific Northwest ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Distribution Functions ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Reactor Pressure

To reduce uncertainties in flaw-related inputs for probabilistic fracture mechanics (PFM) evaluations, the U.S. Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) involving nondestructive and destructive examinations for fabrication flaws in reactor pressure vessel (RPV) material. Using these data, statistical distributions have been developed to characterize the flaws in regions of a RPV. The regions include the main seam welds, repair welds, base metal, and the cladding at the inner surface of the vessel. This paper summarizes the available data and describes the treatment of these data to estimate flaw densities, flaw depth distributions, and flaw aspect ratio distributions. The methodology has generated flaw-related inputs for PFM calculations that have been part of an effort to update pressurized thermal shock (PTS) regulations. Statistical treatments of uncertainties in the parameters of the flaw distribution functions are part of the inputs to the PFM calculations. The paper concludes with a presentation of some example input files that have supported evaluations by USNRC of the risk of vessel failures caused by PTS events.

Sign in / Sign up

Export Citation Format

Share Document