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.

Analysis of Circumferential Welds in BWRs for Life Beyond 60

2015 ◽  
Author(s):  
Patrick Purtscher ◽  
Simon Sheng ◽  
Terry Dickson
Keyword(s):  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Boiling Water Reactors ◽  
Circumferential Welds ◽  
Oak Ridge ◽  
Original Analysis ◽  
Asme Code ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes the probabilistic fracture mechanics (PFM) analyzes of the conditional probability of failure (CPF) due to brittle fracture of circumferential welds (CW) from a cold overpressurize event in boiling water reactors (BWR) operated for 72 EFPY. This analysis used the Fracture Analysis for Vessels, Oak Ridge (FAVOR) computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding. Two typical vessel configurations and the associated material properties for the beltline materials, CW, axial welds (AW), and plates (PL) were used. The analyses consider the potential effects of different fabrication options, shop vs field. Shop-fabrication is mainly by submerged arc weld (SAW) process, while field fabrication used the shielded metal arc weld (SMAW) process. In either case, repairs would have required the SMAW process. The calculations show that field-fabricated vessels would have a slight increase in the CPF compared to shop-fabricated vessels, but the assumed fraction of repair welds was more significant than the fabrication option. The details demonstrate the relative importance of surface-breaking flaws vs. embedded flaws for the assumed transient. The results confirm the conclusions from the original analysis from BWRVIP-05 and BWRVIP-74, the CPF for CW is orders of magnitude less than that of PL and AW regions of the vessel; therefore, the ASME Code-required volumetric examinations of the CW every 10 years as part of the in-service inspection (ISI) program does not change the overall CPF for the vessel. In all the cases analyzed, the total CPF values of the BWRs for 72 EFPY are below the goal for safe operation.

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.

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.

Evaluation of the forced outage length contribution of critical nuclear power systems in an extended operating cycle

2000 ◽  
Vol 214 (1) ◽  
pp. 23-28
Author(s):  
C W Kang
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Nuclear Regulatory Commission ◽  
Capacity Factor ◽  
Plant Performance ◽  
Operating Cycle ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission

The work presented here presents an evaluation method for the question of how reliably the system (or component), responsible for the dominant plant availability loss, will run in an extended 48 month operating cycle. As major contributors to the total plant forced outage time in pressurized water reactors (PWRs), reactor coolant pumps (RCPs) and main feed pumps (MFPs) are chosen as specific example systems for a case study. The method proposed estimates the expected forced outage length contribution of each system to the maximum allowed outage length given a certain plant capacity factor. Based upon the current reliability level estimated from the Nuclear Regulatory Commission plant performance database, the assessment of each system impact shows that 14.2 and 2.2 per cent of the maximum allowed outage length are expected to be taken by RCPs and MFPs respectively in the PWR regardless of other systems. In order to meet a 97 per cent goal capacity factor to be envisaged in a 48 month operating cycle, it is recommended that various possible actions be devised for achieving the higher RCP and MFP operational availability through design, monitoring and maintenance.

A Fuel Cycle and Core Design Analysis Method for New Cladding Acceptance Criteria Using PHISICS, RAVEN and RELAP5-3D

2016 ◽  
Author(s):  
Andrea Alfonsi ◽  
George L. Mesina ◽  
Angelo Zoino ◽  
Cristian Rabiti
Keyword(s):  
Fuel Cycle ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Computational Time ◽  
Maximum Temperature ◽  
Reactor Accident ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Safety Margins ◽  
Regulatory Commission

The Nuclear Regulatory Commission (NRC) has considered revising the 10 CFR 50.46C rule [1] for analyzing reactor accident scenarios to take the effects of burn-up rate into account. Both maximum temperature and oxidation of the cladding must be cast as functions of fuel exposure in order to find limiting conditions, making safety margins dynamic limits that evolve with the operation and reloading of the reactor. In order to perform such new analysis in a reasonable computational time with good accuracy, INL (Idaho National Laboratory) has developed new multi-physics tools by combining existing codes and adding new capabilities. The PHISICS (Parallel Highly Innovative Simulation INL Code System) toolkit [2,3] for neutronic and reactor physics is coupled with RELAP5-3D [4] (Reactor Excursion and Leak Analysis Program) for the LOCA (Loss of Coolant Accident) analysis and RAVEN [5] for the PRA (Probabilistic Risk Assessment) and margin characterization analysis. In order to perform this analysis, the sequence of RELAP5-3D input models had to get executed in a sequence of multiple input decks, each of them had to restart and slightly modify the previous model (in this case, on the neutronic side only) This new RELAP5-3D multi-deck processing capability has application to parameter studies and uncertainty quantification. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical PWR (Pressurized Water Reactor).

Stress-Based Environmental Fatigue Monitoring of PWR Charging Nozzle

2013 ◽  
Author(s):  
Timothy Gilman ◽  
Archana Chinthapalli ◽  
Michael Hoehn
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Technical Report ◽  
Fatigue Monitoring ◽  
Energy Center ◽  
Plant Data ◽  
Water Reactors ◽  
Fatigue Evaluation ◽  
Regulatory Commission

This paper describes the techniques utilized to perform a stress-based environmentally-assisted fatigue evaluation of Westinghouse-designed charging branch nozzles on the reactor coolant loop of the Callaway Energy Center nuclear power plant. Analysis results from using idealized, design transient definitions are compared to those resulting from analysis of the actual plant data. Benchmarking analyses, performed to address Nuclear Regulatory Commission (NRC) concerns about simplified methodologies, are described. The simplified results are also compared to those produced using an advanced, multiaxial stress-based fatigue methodology defined in a recent EPRI technical report [3]. This paper concludes that stress-based fatigue monitoring using actual plant data is an effective way for a plant to manage environmentally-assisted fatigue of charging nozzles in pressurized water reactors (PWRs).

scholarly journals Deployment and Operation of the ES-3100 Type B Shipping Container

2006 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Dean R. Tousley ◽  
Dennis B. Miller
Keyword(s):  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Oak Ridge ◽  
Shipping Container ◽  
Enriched Uranium ◽  
Transportation Research ◽  
Regulatory Commission ◽  
The U.S

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping, for disposition purposes, bulk quantities of fissile materials, primarily highly enriched uranium (HEU). The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material since the 1980s. However, the 6M does not conform to the packaging requirements in the Code of Federal Regulations (10 CFR 71) and, for that reason, is being phased out for use in the DOE secure transportation system by the end of 2006. BWXT Y-12 developed and licensed the ES-3100 container to replace the DOT 6M. The ES-3100 was certified by the Nuclear Regulatory Commission (NRC) in April 2006. The process of deploying the new package began in June 2005 and is planned to be completed in July 2006. The package will be fully operational and completely replace the DOT 6M at the Y-12 National Security Complex (Y-12) by October 2006. This paper reviews the deployment process and the mock loading station that was installed at National Transportation Research Center (NTRC) of Oak Ridge National Laboratory. Specialized equipment, tools, and instrumentation that support the handling and loading operations of the ES-3100 are described in detail. Loading options for other user sites are explored in preparation for deployment of this new state-of-the-art shipping container throughout the DOE complex and the private sector.

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

2009 ◽  
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 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 (LTOP) 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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