Learning Through Delivery, Westinghouse AP1000® Plant Licensing

2014 ◽  
Author(s):  
Julie Gorgemans ◽  
Michael M. Corletti ◽  
Richard A. DeLong ◽  
Terry L. Schulz
Keyword(s):  
South Carolina ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Lessons Learned ◽  
Plant Design ◽  
Pressurized Water ◽  
Design Assessment ◽  
Under Construction ◽  
Safety Features ◽  
The Uk

The AP1000® plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety, and costs. Four AP1000 units are currently under construction on coastal sites of Sanmen and Haiyang, China. Additionally, the United States (US) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) & South Carolina Electric & Gas Company (SCE&G) to construct and operate AP1000 plants at the existing Vogtle & VC Summer sites in Georgia and South Carolina, respectively. Although construction at both US sites is underway, the first four China AP1000 plants will become operational ahead of the U.S. Domestic AP1000 plants. Westinghouse is also actively engaged in deploying the AP1000 plant design in other regions throughout the world such as Europe. For example, the AP1000 plant design was evaluated by the UK Office for Nuclear Regulation as part of the UK Generic Design Assessment and received a statement of interim Design Acceptance in late 2011. This paper reviews the past and on-going AP1000 plant licensing activities and discusses how the significant lessons learned gathered through the AP1000 plant worldwide deployment increase licensing certainty for any new AP1000 project.

Alternate Requirements for Protection Against Pressurized Thermal Shock (PTS)

2013 ◽  
Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Xavier Pitoiset
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Prediction Techniques ◽  
The U.S

Plants in the United States (U.S.) and many plants outside of the U.S. are required to meet the regulations of the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61. The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61). The Alternate PTS Rule was developed through probabilistic fracture mechanics (PFM) evaluations performed for selected U.S. pilot plants. A Generalization Study was also performed which determined that the plants used for these evaluations were representative of and applicable to the U.S. Pressurized Water Reactor (PWR) nuclear power plant fleet. Plants outside of the U.S. may be interested in implementing the Alternate PTS Rule. However, direct implementation of the Alternate PTS Rule may not be possible due to differences in plant design, embrittlement prediction techniques, inservice inspection requirements, etc. The objective of this paper is to explore the use the Alternate PTS Rule by PWR plants outside of the U.S. by proposing methods to account for the potential differences mentioned above.

AP1000® Plant Application of Defense in Depth

2014 ◽  
Author(s):  
Terry L. Schulz ◽  
Julie Gorgemans
Keyword(s):  
United States ◽  
The United States ◽  
Passive Safety ◽  
Plant Design ◽  
Pressurized Water ◽  
Investment Protection ◽  
Regulatory Standards ◽  
Geographical Regions ◽  
Defense In Depth ◽  
Safety Features

The AP1000 plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. One of the key design approaches in the AP1000 plant design is to use passive features to mitigate design basis accidents. Active defense-in-depth (DiD) features provide investment protection, reduce the demands on the passive features and support the Probabilistic Risk Assessment (PRA). The passive features are classified as safety-related in the United States. The active DiD features are classified as nonsafety-related (with supplemental requirements) in the United States. The AP1000 plant design has also incorporated a standardization approach, which together with the level of safety achieved by the passive safety features, results in a plant design that can be applied to different geographical regions with varying regulatory standards and utility expectations without major changes. This paper will discuss the approach taken to defining DiD in the AP1000 plant and the effectiveness of that approach. It will also address the capability of the AP1000 plant to meet deterministic DiD guidelines such as the ones in application in the UK or described in the Western European Nuclear Regulators’ Association (WENRA) safety objectives for new plants.

Update to Risk-Informed Pressurized Water Reactor Vessel 10 to 20 Year Inspection Interval Extension

2006 ◽  
Author(s):  
Nathan A. Palm ◽  
Bruce A. Bishop ◽  
Cheryl L. Boggess
Keyword(s):  
Pilot Plant ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Lessons Learned ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pilot Studies ◽  
Water Reactor ◽  
Inspection Interval ◽  
Interval Extension

The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

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.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

Developments in Structural Integrity Safety Cases for Highest Reliability Components in the UK

2016 ◽  
Author(s):  
T. Jelfs ◽  
M. Hayashi ◽  
A. Toft
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Reliability ◽  
Structural Integrity ◽  
Lessons Learned ◽  
Advisory Group ◽  
Design Assessment ◽  
Safety Case ◽  
The Uk

Gross failure of certain components in nuclear power plant has the potential to lead to intolerable radiological consequences. For these components, UK regulatory expectations require that the probability of gross failure must be shown to be so low that it can be discounted, i.e. that it is incredible. For prospective vendors of nuclear power plant in the UK, with established designs, the demonstration of “incredibility of failure” can be an onerous requirement carrying a high burden of proof. Requesting parties may need to commit to supplementary manufacturing inspection, augmented material testing requirements, enhanced defect tolerance assessment, enhanced material specifications or even changes to design and manufacturing processes. A key part of this demonstration is the presentation of the structural integrity safety case argument. UK practice is to develop a safety case that incorporates the notion of ‘conceptual defence-in-depth’ to demonstrate the highest structural reliability. In support of recent Generic Design Assessment (GDA) submissions, significant experience has been gained in the development of so called “incredibility of failure” arguments. This paper presents an overview of some of the lessons learned relating to the identification of the highest reliability components, the development of the structural integrity safety arguments in the context of current GDA projects, and considers how the UK Technical Advisory Group on Structural Integrity (TAGSI) recommendations continue to be applied almost 15 years after their work was first published. The paper also reports the approach adopted by Horizon Nuclear Power and their partners to develop the structural integrity safety case in support of the GDA process to build the UK’s first commercial Boiling Water Reactor design.

LBB Under Beyond Design Basis Seismic Loading

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Heavy Water Reactor ◽  
Regulatory Commission

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

Production of custom-made recycled water for various reuse purposes: lessons learned from one of the world's largest recycling facilities

2012 ◽  
Vol 7 (3) ◽  
Author(s):  
Valentina Lazarova ◽  
Gregg Oelker ◽  
Wyatt Oelker
Keyword(s):  
Water Quality ◽  
Leading Edge ◽  
The United States ◽  
Lessons Learned ◽  
Successful Implementation ◽  
Plant Design ◽  
Recycled Water ◽  
Integrated Resource Management ◽  
Custom Made ◽  
The World

A new concept developed for the successful implementation of integrated resource management in urban and protected areas is the production of ‘designer’ or custom-made recycled water for various reuse purposes. The Edward C. Little Water Recycling Facility in El Segundo, California is an excellent example and the only facility in the world and the United States producing five distinct types of recycled water. This paper presents the plant design, critical milestones, water quality challenges faced and problems solved towards consistently meeting the needs of a diverse client base through delivery of a leading-edge range of custom-made recycled waters.

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