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.

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 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.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Various MTC Studies of TRACE with RCS Pressure Predictions under ATWS for Maanshan

2013 ◽  
Vol 284-287 ◽  
pp. 1151-1155
Author(s):  
Che Hao Chen ◽  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Graphic User Interface ◽  
Pressurized Water ◽  
Plant Safety ◽  
Regulatory Commission

The objective of this study is to utilize TRACE (TRAC/RELAP Advanced Computational Engine) code to analyze the reactor coolant system (RCS) pressure transients under ATWS (Anticipated Transient Without Scram) for Maanshan PWR (Pressurized Water Reactor) in various MTC (Moderator Temperature Coefficient) conditions. TRACE is an advanced thermal hydraulic code for nuclear power plant safety analysis, which is currently under development by the United States Nuclear Regulatory Commission (USNRC). A graphic user interface program named SNAP (Symbolic Nuclear Analysis Package), which processes inputs and outputs for TRACE is also under development. Maanshan nuclear power plant (NPP) is the only Westinghouse PWR in Taiwan. The rated core thermal power of Maanshan with MUR (Measurement Uncertainty Recapture) is 2822 MWt. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. According to ASME Code Level C service limit criteria, the RCS pressure for Maanshan NPP must be under 22.06 MPa. Furthermore, we select the LOCV transient to analyze various MTC effects on RCS pressure variations.

scholarly journals Regulatory Treatment of Nonsafety Systems in Passive Advanced Light-Water Reactors

2014 ◽  
Author(s):  
Thomas Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Active Systems ◽  
Light Water ◽  
Defense In Depth ◽  
Water Reactors ◽  
Regulatory Commission ◽  
High Level

Some new nuclear power plants have advanced light-water reactor (ALWR) designs with passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy, to supply safety-injection water and to provide reactor-core and containment cooling. Active systems in such passive ALWR designs are categorized as nonsafety systems with limited exceptions. Active systems in passive ALWR designs provide the first line of defense to reduce challenges to the passive systems in the event of a transient at the nuclear power plant. Active systems that provide a defense-in-depth function in passive ALWR designs need not meet all of the acceptance criteria for safety-related systems. However, there should be a high level of confidence that these active systems will be available and reliable when challenged. Multiple activities will provide confidence in the capability of these active systems to perform their defense-in-depth functions; these are collectively referred to as the Regulatory Treatment of Nonsafety Systems (RTNSS) program. The U.S. Nuclear Regulatory Commission (NRC) addresses policy and technical issues associated with RTNSS equipment in passive ALWRs in several documents. This paper discusses the NRC staff’s review of pumps, valves, and dynamic restraints within the scope of the RTNSS program in passive ALWRs. Paper published with permission.

Technology Lecture

1983 ◽  
Vol 385 (1789) ◽  
pp. 241-251 ◽  
Keyword(s):  
Nuclear Power ◽  
Task Force ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Accidents ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Turbine Generators ◽  
Set Up ◽  
Regulatory Commission

The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the U.K. the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear powrer focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The vocabulary used in the nuclear power industry contributes to the misunderstanding and fear felt by the general public. The long-term consequences of big nuclear accidents can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking.

Results of Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model

2002 ◽  
Author(s):  
Michael F. Hessheimer ◽  
Satoru Shibata ◽  
James F. Costello
Keyword(s):  
Prestressed Concrete ◽  
Nuclear Power ◽  
Limit State ◽  
Nuclear Regulatory Commission ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Containment Vessel ◽  
Engineering Corporation ◽  
Regulatory Commission

The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC) have been co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories. The purpose of the program is to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions with measured behavior. This is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. The first project in this program was a test of a mixed scale steel containment vessel (SCV). Next, a 1:4-scale model of a prestressed concrete containment vessel (PCCV), representative of a pressurized water reactor (PWR) plant in Japan, was constructed by NUPEC at Sandia National Laboratories from January 1997 through June, 2000. Concurrently, Sandia instrumented the model with over 1500 transducers to measure strain, displacement and forces in the model from prestressing through the pressure testing. The limit state test of the PCCV model was conducted in September, 2000 at Sandia National Laboratories. This paper describes the conduct and some of the results of this test.

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.

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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