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Fleming K.N.,KNF Consulting Services LLC | Lydell B.O.Y.,Scandpower Risk Management | Grantom C.R.,STP Nuclear Operating Company
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2012

This paper summarizes the analysis of loss of coolant accident (LOCA) frequencies in support of a risk-informed (Ri) evaluation of Generic Safety issue (GSi) 191 for the South Texas Project Electric Generating Station (STPEGS) Units 1 and 2. The STP Ri-GSi 191 Closure Study investigates the size and location of loss of coolant accidents (LOCAs) more finely than in traditional PRAs in order to assess the risk of debris formation during the LOCAs that could interfere with the operation of the emergency core cooling systems (ECCSs) or inhibit coolant flow through the core during the recirculation phase after a LOCA. The size and location of the break could influence the amount and chemistry of debris formation and the timing and need for actions to initiate or terminate containment sprays and recirculation cooling. This application requires the capability to estimate LOCA frequencies as a function of break size at each location within the Class 1 pressure boundary with due regard to the proper characterization and quantification of uncertainties. Copyright © 2012 by ASME. Source


Kocher J.A.,Conco Services Corporation | Frazee R.,STP Nuclear Operating Company | Wolf M.,2800 Louis Lumiere
American Society of Mechanical Engineers, Power Division (Publication) POWER | Year: 2014

Eddy Current Testing (ECT) of condenser tubes is essential to maintaining good plant reliability and availability. Early identification of defects can allow for adequate remedial action and prevent forced outages caused by condenser tube leaks. The well-known catastrophic failure in the nuclear industry in Japan has not only raised concern in Japan over aging nuclear power plants, but has also raised concern over safe operations in the United States and around the world. Ongoing reliability and instability issues due to reported leaks in condensers have also been the topic for nuclear watchdogs. This focus on the nuclear plant condenser has brought to light the various levels of sophistication and capability in ECT. In ECT, the type of defect present in a condenser tube is determined by the characteristics it presents under test. The tubes must be adequately cleaned prior to testing and some awareness or evidence of the type of defect to be uncovered should be available to the testing team. In cases where defects are discovered that are inconsistent with prior awareness further exploratory testing is common. Exploratory testing can proceed to test areas of suspected defects in the tubing, and it may result in a complete redefinition of the test procedure, inclusive of instruments, probe types and other key ECT criteria. A comprehensive knowledge of testing options and their practical application is necessary to redefine a test that will yield meaningful results and achieve the intended objective; to identify the type and extent of defect and take remedial action therefore preventing failure. This paper addresses such a case at the South Texas Project (STP) Nuclear Power Plant where peculiar defects were undeterminable under standard ECT procedures. The defects continued to negatively impact reliability and stability at the plant until a new ECT process and test procedure were developed, demonstrated and deployed. The result achieved was accurate defect detectability and improved nuclear plant reliability. Copyright © 2014 by ASME. Source


Liming J.K.,ABSG Consulting | Grantom C.R.,STP Nuclear Operating Company
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013 | Year: 2013

This paper summarizes recent lessons learned regarding risk-informed surveillance frequency control program (RI-SFCP) implementation at commercial nuclear power stations. Since 2008, the authors of this paper have played significant roles in implementing industry initiative 5b RI-SFCPs for 20 nuclear power generating units operated by eight nuclear power utility companies. These programs include development, review, and implementation of surveillance test risk-informed documented evaluation (STRIDE) packages designed to support extension of conventional surveillance requirement test intervals, in accordance with "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guideline" (NEI 04-10, Revision 1). The scope of work associated with STRIDE development and implementation, which the authors have supported, includes probabilistic risk assessment (PRA) case studies, deterministic assessment (DA), and, where required, instrument drift evaluation (IDE). The STRIDE implementation efforts have also included support of independent decision-making panel (IDP) meetings at the implementing power stations. The purpose of this paper is to provide a presentation of lessons learned during the considerable STRIDE development and implementation experience of the authors, which includes support for the development and implementation of over 80 plant STRIDEs. The major focus of the paper is lessons learned associated with STRIDE PRA case study development and implementation, but also includes insights about associated STRIDE DA and IDE development and implementation. The scope of the discussion in this paper includes treatment of conventional deterministic safety criteria as well as probabilistic risk criteria. The paper addresses both qualitative and quantitative aspects relating to STRIDE implementation. Source


Liming J.K.,ABSG Consulting | Johnson D.H.,ABSG Consulting | Grantom C.R.,STP Nuclear Operating Company
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011 | Year: 2011

This paper summarizes a refreshed perspective on a proposed integrated risk-informed performance-based regulatory framework via the application of probabilistic safety assessment (PSA). This perspective is refreshed, in that it is based on the considerable industry experience gained during the last decade in the implementation of important risk-informed applications (e.g., risk-managed technical specifications (RMTS), risk-informed surveillance frequency control programs (RI-SFCPs), risk-informed in-service testing programs (RI-IST), risk-informed in-service inspection (RI-ISI) programs, risk-informed graded quality assurance (RI-GQA) programs, etc.) and in the area of PSA standards development and implementation. The focus of this paper is to provide an integrated framework of proposed practical safety management metrics that can be effectively and efficiently applied in the regulation of commercial nuclear power plant design, construction, operation, maintenance, and decommissioning. The scope of the discussion in this paper includes treatment of conventional deterministic safety criteria as well as probabilistic risk criteria. The paper addresses both qualitative and quantitative aspects relating to this proposed regulatory framework. Source


Liming J.K.,ABSG Consulting | Mikschl T.J.,ABSG Consulting | Rodgers S.S.,STP Nuclear Operating Company
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011 | Year: 2011

This paper summarizes the results of an evaluation of human action dependency for the STP Nuclear Operating Company (STPNOC) South Texas Project Electric Generating Station (STPEGS) Units 1 and 2 low power and shutdown (LPSD) probabilistic risk assessment (PRA). Specifically, this paper focuses on the potential impact of refinements to current industry PRA human reliability analysis (HRA) methods (e.g., the EPRI HRA Calculator® methods) for human action dependency evaluation. These potential refinements were conceptualized during the performance of the STPNOC LPSD PRA HRA. The scope of this evaluation included a thorough post-processing evaluation of over 37,000 PRA event sequences (or cut sets) for combinations of human failure events (HFEs) that could result in potential HEP interdependence, and thus, could significantly impact the results of the PRA and any associated risk-informed applications. The paper presents a discussion of the importance of human action dependency analysis (HADA) in PRA or probabilistic safety assessment (PSA), and presents an overview of current methods typically applied. The paper also presents general results from the STPNOC LPSD PRA HRA HADA, and it provides selected examples of how potential HADA refinements could impact the rigor and accuracy of HADA results, and thus, overall PRA or PSA results. Source

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