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Liming J.K.,ABSG Consulting | Quinn E.L.,Technology Resources
9th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2015 | Year: 2015

This paper summarizes an updated process for risk-informed surveillance frequency control program (RI-SFCP) implementation at nuclear power stations based on lessons learned in recent years. Since 2008, the authors of this paper have played significant roles in implementing industry initiative 5b RI-SFCPs for 20 nuclear power units operated by 8 nuclear power utility companies. These programs include selection and prioritization of specific target surveillance test interval extensions; and 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 [1]. The scope of work associated with STRIDE development includes probabilistic risk assessment (PRA) case studies, deterministic assessment (DA) evaluations, 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 and IDP member training. The purpose of this paper is to provide a presentation of a refined process for STRIDE development with a focus on instrumentation and control systems based on author experience, which includes support for the development of over 100 plant STRIDES. This paper outlines a framework for practical implementation of an RI-SFCP within the context of an integrated risk-informed performance-based regulation application program with emphasis on instrumentation and control systems. Source


Schrader K.J.,Pacific Gas and Electric Company | Hefler J.W.,Altran GmbH | Quinn E.L.,Technology Resources
9th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2015 | Year: 2015

Diablo Canyon Power Plant (DCPP) is replacing the existing digital Westinghouse Eagle 21 Process Protection System (PPS) to address obsolescence issues. Eagle 21 was installed in 1994 to replace the original analog Westinghouse 7100 PPS. The PPS replacement design is based on a combination of the Invensys Tricon VI0 programmable logic controller and the Westinghouse Advanced Logic System field programmable gate array digital instrumentation and control devices. The License Amendment for replacement of the Eagle 21 PPS was submitted to the NRC on October 26, 2011. Key to submittal of the PPS replacement LAR was resolution of the need for Diversity and Defense-in-Depth (D3) in the replacement design to mitigate the potential for a common design error to disable redundant channels of the protection systems through common-cause failure (CCF). The PG&E PPS Replacement Project D3 Assessment Topical Report was submitted to the NRC in April, 2010, and approved in April, 2011. This paper discusses the architecture of the PPS replacement design and the methodology by which PG&E assessed the diversity requirements of the Diablo Canyon Power Plant (DCPP) digital PPS relative to current regulations and guidance, and developed a design with sufficient built-in diversity to meet USNRC DI&C ISG-02 Staff Position 1 without a Diverse Actuation System (DAS). Source


Smidts C.,Ohio State University | Huang F.,Ohio State University | Li X.,Ohio State University | Mutha C.,Ohio State University | Quinn T.,Technology Resources
9th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2015 | Year: 2015

The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation and control systems in safety critical applications has shown itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability attributes include reliability, safety, availability, maintainability, and security (confidentiality and integrity). Modeling the dependencies between the dependability attributes is the first step towards dependability quantification. In this research we use two methods: structured expert opinion elicitation and (hierarchical) causal mapping to extract the dependencies. A panel of fourteen international experts was identified. Each expert filled a unique questionnaire, targeted towards dependability and attributes as per his/her expertise. The questionnaires were designed in a semi-structured format. The questions were designed to elicit the attributes encompassed by dependability, the root causes of each attribute, the dependencies between attributes, and how root causes and attributes affect dependability. Then the data from the expert elicitation was analyzed and converted to fourteen hierarchical causal maps. A hierarchical causal map is divided into three levels of detail: the top layer of the causal map is called the dependence level composed of the dependability attributes and interrelationships; the middle layer is called the Event of interest (Eol) level and expresses mechanisms leading to occurrence of the main event of interest (for instance a safety critical failure) for each dependability attribute; the third layer is called Measureable Concepts level, and is composed of measures for each of the Eol contributors. Finally, a merged causal map on the dependencies between dependability attributes was developed. Source


Schrader K.J.,Pacific Gas and Electric Company | Hefler J.W.,Altran GmbH | Quinn E.L.,Technology Resources
9th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2015 | Year: 2015

This paper provides a summary of the preparation of a License Amendment Request (LAR) for replacement of the existing Diablo Canyon Power Plant (DCPP) Eagle 21 digital Process Protection System (PPS) by Pacific Gas & Electric Company (PG&E) using US Nuclear Regulatory Commission (USNRC) Digital I&C (DI&C) Interim Staff Guidance (ISG) document ISG-06, Licensing Process [1]. The USNRC has designated the LAR for the DCPP PPS replacement project as the pilot application for use of DI&C ISG-06. DI&C ISG-06, Revision 1, issued for use on January 19, 2011, describes the licensing process the NRC staff may use to review an LAR for a DI&C modification. The LAR [2] for the replacement of the Eagle 21 PPS was submitted to the NRC on October 26, 2011, was accepted for review [3] on January 13, 2012, and is currently under review with approval expected in late 2015. A more detailed review of the specific PPS replacement design including the Defense-In-Depth and Diversity approach is provided in another paper being presented at this conference [12]. The use of pre-application (Phase 0) public meetings with the NRC staff to ensure the PPS replacement design adequately addressed NRC criteria was instrumental to development of the proposed PPS replacement design. This paper discusses the topics discussed in the Phase 0 meetings and the approach utilized by PG&E for the Phase 0 public meetings to support efficient development of the proposed PPS replacement design. PG&E developed a PPS Replacement diversity and defense-in-depth (D3) Assessment Topical Report [4] that was submitted to NRC in 2010 and approved [5] in 2011. This paper discusses the PG&E use of the D3 assessment to evaluate and optimize the initial PPS replacement design prior to submittal of the design to the NRC for approval. The PG&E experience with preparation of the LAR contents and supporting documents and lessons learned during the detailed design and verification phase is discussed. Source


Pietre-Cambacedes L.,Electricite de France | Quinn E.L.,Technology Resources | Hardin L.,U.S. Nuclear Regulatory Commission
IFAC Proceedings Volumes (IFAC-PapersOnline) | Year: 2013

This paper provides an overview of the work of the International Electrotechnical Commission (IEC) on the development of a series of standards dealing with the cyber security of nuclear power plant (NPP) instrumentation and control (I&C) systems. In particular, the status and content of the first, top level document of the series, IEC 62645, is described. A more recent draft, IEC 62859, dealing with the coordination between safety and cyber security aspects, is also presented. Future work and perspectives associated with this new series of standards are finally discussed. © IFAC. Source

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