Regulatory Perspective for the Definition of Probabilistic Acceptance Criteria for CANDU Pressure Tubes

2016 ◽  
Author(s):  
Sankar Laxman ◽  
Blair Carroll ◽  
John Jin
Keyword(s):  
Probabilistic Method ◽  
Operating Experience ◽  
Service Evaluation ◽  
Steering Committee ◽  
Canadian Standard Association ◽  
Degradation Mechanisms ◽  
Acceptance Criteria ◽  
Reactor Unit ◽  
Technical Requirements ◽  
Pressure Tubes

For the assurance of fitness-for-service of CANDU Pressure Tubes (PTs), guidelines and acceptance criteria are provided in Canadian Standard Association (CSA) N285.8-15, Technical requirements for in-service evaluation of zirconium alloy pressure Tubes in CANDU reactors. With respect to the assessment of risk of operation associated with degradation mechanisms and aging of the PTs in the entire core of a given reactor Unit, Clause 7 of CSA N285.8 allows Licensee’s to use either a deterministic or probabilistic method to assess the likelihood of PT failures. When a probabilistic method is used, the Licensee is obligated to demonstrate that the combined frequency of PT failure(s) over the evaluation period, due to the various potential degradation mechanisms, is less than the maximum acceptable frequency provided in Table C.1 of CSA N285.8-15. The maximum acceptable frequency provided in Table C.1 of CSA N285.8-15 was developed in the early-1990’s based on reactor operating experience and knowledge at that time, Station Siting Guides and Consultative Regulatory Guide C-006 (Revision 1). A task group was established by the CSA N285.8 Technical Steering Committee to re-evaluate the allowable failure frequencies to confirm that they remain relevant given the current state of knowledge and the additional evaluation tools available. This paper provides Canadian Nuclear Safety Commission staff views regarding the technical basis for revisions to the allowable frequencies based upon current industry practices in conducting probabilistic core assessments.

Probabilistic Assessment of CANDU Reactor Core for Risk of Pressure Tube Failure due to Presence of In-Service Flaws

2017 ◽  
Author(s):  
David Cho ◽  
Danny H. B. Mok ◽  
Steven X. Xu ◽  
Douglas A. Scarth
Keyword(s):  
Crack Initiation ◽  
Crack Growth ◽  
Reactor Core ◽  
Pressure Tube ◽  
Probabilistic Assessment ◽  
Degradation Mechanisms ◽  
Technical Requirements ◽  
Fuel Bundle ◽  
Pressure Tubes ◽  
Tube Failure

Technical requirements for analytical evaluation of in-service Zr-2.5Nb pressure tubes in CANDU(1) reactors are provided in the Canadian Standards Associate (CSA) N285.8. The evaluation must address all in-service degradation mechanisms including the presence of in-service flaws. Flaws found during in-service inspection of CANDU Zr-2.5Nb pressure tubes, including fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws, dummy bundle bearing pad fretting flaws, erosion-shot flaws and crevice corrosion flaws, are volumetric and blunt in nature. These in-service flaws can become crack initiation sites during pressure tube operation and potentially lead to pressure tube failure. Any detected flaws that do not satisfy the criteria of acceptance as per CSA N285.4 must be analytically evaluated to justify continued operation of the pressure tube. Moreover, the risk of pressure tube failure due to presence of in-service flaws in the entire reactor core must be assessed. A review of assessment of the risk of pressure tube failure due to presence of in-service flaws in CANDU reactor core is provided in this paper. The review covers the technical requirements in the CSA N285.8 for evaluating degradation mechanisms related to flaws in the reactor core. Current Canadian industry practice of probabilistic assessment of reactor core for pressure tube failure due to presence of in-service flaws is described, including evaluation of flaws for crack initiation, subsequent crack growth to through-wall penetration, and pressure tube rupture due to unstable crack growth prior to safe shutdown of the reactor. Operating experience with the application of probabilistic assessment of reactor core for the risk of pressure tube failure due to presence of in-service flaws is also provided.

A Risk-Informed Approach to Leak-Before-Break Assessment of Pressure Tubes in CANDU Reactors

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
M. D. Pandey ◽  
A. K. Sahoo
Keyword(s):  
Computation Time ◽  
Service Evaluation ◽  
Probabilistic Methods ◽  
Pressure Tube ◽  
Deterministic Method ◽  
Deterministic Analysis ◽  
Candu Reactors ◽  
Technical Requirements ◽  
Pressure Tubes ◽  
Leak Before Break

The leak-before-break (LBB) assessment of pressure tubes is intended to demonstrate that in the event of through-wall cracking of the tube, there will be sufficient time followed by the leak detection, for a controlled shutdown of the reactor prior to the rupture of the pressure tube. CSA Standard N285.8 (2005, “Technical Requirements for In-Service Evaluation of Zirconium Alloy Pressure Tubes in CANDU Reactors,” Canadian Standards Association) has specified deterministic and probabilistic methods for LBB assessment. Although the deterministic method is simple, the associated degree of conservatism is not quantified and it does not provide a risk-informed basis for the fitness for service assessment. On the other hand, full probabilistic methods based on simulations require excessive amount of information and computation time, making them impractical for routine LBB assessment work. This paper presents an innovative, semiprobabilistic method that bridges the gap between a simple deterministic analysis and complex simulations. In the proposed method, a deterministic criterion of CSA Standard N285.8 is calibrated to specified target probabilities of pressure tube rupture based on the concept of partial factors. This paper also highlights the conservatism associated with the current CSA Standard. The main advantage of the proposed approach is that it retains the simplicity of the deterministic method, yet it provides a practical, risk-informed basis for LBB assessment.

COG Risk-Informed In-Service Inspection (RI-ISI) Project

2012 ◽  
Author(s):  
Patrick J. O’Regan ◽  
Mehdi Rezaie-Manesh ◽  
Scott T. Chesworth
Keyword(s):  
Operating Experience ◽  
Nuclear Safety ◽  
Degradation Mechanisms ◽  
Acceptance Criteria ◽  
Risk Levels ◽  
Periodic Inspection ◽  
Starting Point ◽  
Best Fit ◽  
Service Inspection

In 2009, the Canadian Nuclear Safety Commission (CNSC) prepared “Guidelines for Risk-Informed In-service Inspection for Piping,” which provided guidance and acceptance criteria for licensees to develop an RI-ISI program as an alternate to the current CSA N285.4 and augmented programs for piping inspection. A project was initiated by the CANDU Owners Group (COG) to develop a best-fit RI-ISI methodology for the CANDU design and evaluate plant risk levels between current and RI-ISI inspection programs. The traditional EPRI RI-ISI methodology was selected as a starting point, and four plant systems were evaluated. Both failure potential and consequence of failure were used to establish the risk significance for all in-scope piping components. Once the risk associated with each component was established, elements were selected for inspection based upon the sampling percentages of the EPRI RI-ISI methodology, and as a final check a comparison was made between plant risk under the current CSA N285.4 (and augmented) inspection program and under RI-ISI. The project was successfully concluded in 2011, and results confirmed that the EPRI RI-ISI methodology can be adapted to the CANDU design and the degradation mechanisms evaluated under RI-ISI are consistent with CANDU operating experience. The original CSA N285.4 basis for the CANDU Periodic Inspection Program (PIP) was validated, and potential improvements to the station inspection programs were identified.

Inconel X-750 Spacer Fitness-for-Service

2019 ◽  
Author(s):  
Winnie Lau ◽  
Douglas Scarth ◽  
Preeti Doddihal
Keyword(s):  
Reactor Core ◽  
Later Life ◽  
Service Evaluation ◽  
Surveillance Program ◽  
Material Degradation ◽  
Load Carrying ◽  
Gas System ◽  
Pressure Tubes ◽  
Endurance Tests ◽  
Design Requirements

Abstract The CANDU1 (CANada Deuterium Uranium) reactor core consists of 380–480 horizontal Zr-Nb pressure tubes, which support fuel bundles and provide pressurized heavy water cooling. The pressure tubes are supported by fuel channel annulus spacers, which maintain the gap between the hot pressure tube and colder calandria tube while providing a means of leak detection through the annulus gas system. Research and testing in this area have shown that spacer material degradation in later life operation could impact the ability of the component to meet its design requirements. This paper presents a fitness-for-service strategy that could be utilized in demonstrating continued safe operation of these components. Fitness-for-service is based on analysis of crush tests on ex-service spacers to determine the load carrying capacity projected into the future and endurance tests to determine fatigue life. This paper describes these technical approaches and their application in fitness-for-service evaluation of spacers in CANDU operating plants to satisfy requirements for an annulus spacer surveillance program under Clause 12.5 of the CSA Standard N285.4-14.

Building the Technical Case for Long-Term Operation

2015 ◽  
Author(s):  
Richard Tilley ◽  
Robin Dyle
Keyword(s):  
Operating Experience ◽  
Academic Community ◽  
Degradation Mechanisms ◽  
Comprehensive Understanding ◽  
Term Operation ◽  
Societal Benefits ◽  
Actual Operating ◽  
Nuclear Plants ◽  
The Us

United States (US) and International utilities are actively engaged in assessing the economic and societal benefits of operating nuclear plants beyond their initial license periods. Nuclear plant generated electricity is still the largest contributor to non-carbon dioxide emitting generation. In the US, a majority of operating plants has already received approval for an additional 20 years of operation, and soon it is expected that utilities will begin the process to seek a second 20 year renewal. The keys to successful renewal are to maintain safe and reliable operations by building a sound technical case through the following activities: • Develop comprehensive understanding of aging degradation issues for systems, structures and components (SSCs) • Implement specific plant aging management programs to address aging degradation • Confirm behavior of degradation mechanisms for the entire period of operation This paper will step through the above elements to illustrate how a strong technical case may be created for safe and reliable long-term operation. Examples or case studies will be provided to clearly link the fundamental science of materials degradation to the inspection, testing and evaluation efforts implemented at a plant and to the confirmatory data that is provided by both actual operating experience and the extensive research and development projects pursued by industry, governments, and the academic community.

Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation

2000 ◽  
Vol 123 (1) ◽  
pp. 58-64 ◽  
Author(s):  
Fredric A. Simonen ◽  
Stephen R. Gosselin
Keyword(s):  
Life Prediction ◽  
Nuclear Power ◽  
Fatigue Crack Initiation ◽  
Operating Experience ◽  
Degradation Mechanisms ◽  
Original Design ◽  
Structural Degradation ◽  
Plant Safety ◽  
Degradation Data ◽  
Design Life

This paper describes industry programs to manage structural degradation and to justify continued operation of nuclear components when unexpected degradation has been encountered due to design materials and/or operational problems. Other issues have been related to operation of components beyond their original design life in cases where there is no evidence of fatigue crack initiation or other forms of structural degradation. Data from plant operating experience have been applied in combination with inservice inspections and degradation management programs to ensure that the degradation mechanisms do not adversely impact plant safety. Probabilistic fracture mechanics calculations are presented to demonstrate how component failure probabilities can be managed through augmented inservice inspection programs.

scholarly journals Development of the WR-21 Gas Turbine Recuperator

1999 ◽  
Author(s):  
Robert C. Sanders ◽  
George C. Louie
Keyword(s):  
Gas Turbine ◽  
Failure Modes ◽  
Waste Heat ◽  
Operating Experience ◽  
Gas Turbine Engine ◽  
Analytical Models ◽  
Royal Navy ◽  
Engine Exhaust ◽  
Turbine Engine ◽  
Technical Requirements

WR-21 is an intercooled and recuperated (ICR) gas turbine engine being developed by the U. S. Navy (USN) with contributions from the Royal Navy and the French Navy. A key component of the WR-21 engine is the recuperator used to recover waste heat from engine exhaust gas. The recuperator is being designed and fabricated by AlliedSignal Aerospace Company under subcontract to Northrop Grumman Marine Services, the prime contractor for the WR-21 gas turbine engine. One of the most challenging developmental items for the WR-21 engine has proven to be the recuperator. This paper discusses the development of the recuperator, including the advanced development (AD) recuperator which failed after a few hours of operation, the limited operating unit (LOU) recuperator which has supported much of the WR-21 engine development testing and the engineering development model (EDM) recuperator which will be used for a 3000 hour engine endurance test. Included is an overview of USN technical requirements for the recuperator and a review of operating experience with the AD and LOU recuperators. Failure modes that have been experienced are discussed in detail, including root cause evaluations and design modifications. Steps taken to extend the life of the LOU recuperator are discussed. In addition, testing (both single core and full size recuperator) and analytical models that have been used to improve the design and reliability of the recuperator are addressed.

Acceptance Criteria for Probabilistic Fracture Protection Evaluations of CANDU Zr-Nb Pressure Tubes

2018 ◽  
Author(s):  
Douglas Scarth ◽  
Leonid Gutkin
Keyword(s):  
Fracture Toughness ◽  
Conditional Probability ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Service Level ◽  
Pressure Tube ◽  
Stability Evaluation ◽  
Acceptance Criteria ◽  
Pressure Tubes ◽  
Technical Basis

Requirements for pressure-temperature limits to protect against rupture of CANDU nuclear reactor Zr-Nb pressure tubes are provided in the Canadian Standards Association (CSA) Standard N285.8. The requirements are based on a stability evaluation of a postulated axial through-wall flaw for all ASME Service Level A, B, C and D loadings. The flaw stability evaluation is strongly dependent on the fracture toughness of the Zr-Nb pressure tube material. The fracture toughness of Zr-Nb pressure tubes is decreasing with operating hours. The decrease in fracture toughness as well as compounding conservatisms based on using bounding values make deterministic evaluations more challenging. The CSA Standard N285.8 permits probabilistic evaluations of fracture protection, but does not provide acceptance criteria. Proposed acceptance criteria that meet the intent of the design basis for Zr-Nb pressure tubes have been developed. The proposed acceptance criteria consist of a proposed maximum allowable conditional probability of pressure tube rupture for the entire reactor core, as well as a proposed maximum allowable conditional probability of rupture of a single pressure tube. The paper provides a description of the technical basis for the proposed acceptance criteria for probabilistic evaluations of fracture protection.

Development of Evaluation Procedures for Crack Initiation From In-Service Flaws in CANDU Zr-2.5Nb Pressure Tubes With Hydrogen

2016 ◽  
Author(s):  
David Cho ◽  
Steven X. Xu ◽  
Douglas A. Scarth ◽  
Gordon K. Shek
Keyword(s):  
Crack Initiation ◽  
Hydrogen Isotope ◽  
Operating Conditions ◽  
Pressure Tube ◽  
Evaluation Procedure ◽  
Operating Pressure ◽  
Evaluation Procedures ◽  
Technical Requirements ◽  
Fuel Bundle ◽  
Pressure Tubes

Flaws found during in-service inspection of CANDU(1) Zr-2.5Nb pressure tubes include fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws and crevice corrosion flaws. These flaws are volumetric and blunt in nature. Crack initiation from in-service flaws can be caused by the presence of hydrogen in operating pressure tubes and resultant formation of hydrided regions at the flaw tips during reactor heat-up and cool-down cycles. Zr-2.5Nb pressure tubes in the as-manufactured condition contain hydrogen as an impurity element. During operation, the pressure tube absorbs deuterium, which is a hydrogen isotope, from the corrosion reaction of the zirconium with the heavy water coolant. In addition, deuterium ingresses into the pressure tube in the rolled joint region. The level of hydrogen isotope in pressure tubes increases with operating time. Over the years, Canadian CANDU industry has carried out extensive experimental and analytical programs to develop evaluation procedures for crack initiation from in-service flaws in Zr-2.5Nb pressure tubes. Crack initiation experiments were performed on pressure tube specimens with machined notches to quantify resistance to crack initiation under various simulated flaw geometries and operating conditions such as operating load and hydrogen concentration. Predictive engineering models for crack initiation have been developed based on understandings of crack initiation and experimental data. A set of technical requirements, including engineering procedures and acceptance criteria, for evaluation of crack initiation from in-service flaws in operating pressure tubes has been developed and implemented in the CSA Standard N285.8. A high level review of the development of these flaw evaluation procedures is described in this paper. Operating experience with the application of the developed flaw evaluation procedure is also provided.

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