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.

A Framework for Estimating Burst Test Fracture Toughness for Zr-2.5Nb Pressure Tubes Using Data From Small Specimen Tests

2019 ◽  
Author(s):  
Steven X. Xu ◽  
Kim Wallin ◽  
David Cho
Keyword(s):  
Fracture Toughness ◽  
Low Cost ◽  
Pressure Tube ◽  
Current Status ◽  
Operating Pressure ◽  
Small Specimen ◽  
Burst Test ◽  
Pressure Tubes ◽  
Using Data ◽  
Technical Basis

Abstract Zr-2.5Nb pressure tubes are primary pressure boundaries in a CANDU2 reactor. Design of pressure tube dimensions allows testing of a pressure tube section at its full size in the laboratory. Burst tests, i.e., internally pressuring pressure tube sections containing axial through-wall cracks till burst, have been used to provide test data of fracture toughness for pressure tubes with axial flaws. The advantage of measuring fracture toughness from burst tests is that measured toughness values are directly applicable to operating pressure tubes. Burst tests, however, are costly and consume considerable amount of material. Only a small number of burst tests can be performed in practice. There is strong motivation to estimate burst test fracture toughness using data from small specimen tests. The estimated burst test fracture toughness can fill the gap in the measured burst test toughness data, as well as provide information on material variability and data scatter. The technical challenge for estimating burst test toughness is that the estimated burst test toughness using data from low cost, small specimen tests must be reliable and representative of burst test specimen behavior with high confidence. A framework for accurately estimating burst test toughness using data from curved compact tests has been under development and is described in this paper. Aspects of technical basis and current status of developing analytical procedures for systematically estimating burst test toughness are presented.

Probabilistic Integrity Assessment of Axial Flaw in CANDU Pressure Tube Due to the Diameter and Thickness Variation

2002 ◽  
Author(s):  
Sang-Log Kwak ◽  
Joon-Seong Lee ◽  
Young-Jin Kim ◽  
Youn-Won Park
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Assessment Method ◽  
Pressure Tube ◽  
Thickness Variation ◽  
Initial Surface ◽  
Size Change ◽  
Integrity Assessment ◽  
Pressure Tubes ◽  
Failure Probabilities

In the CANDU nuclear reactor, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and heavy water coolant. The pressure tubes are installed horizontally inside the reactors and only selected samples are periodically examined during In-Service Inspection (ISI) due to numerous numbers of tubes. Also, these tubes gradually pick up deuterium, as such are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. Up to this time, integrity evaluations are performed using conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. This paper describes probabilistic fracture mechanics analyses of the pressure tubes in consideration of the diameter and thickness variation. Initial hydrogen concentration, the depth and aspect ratio of an initial surface crack, the DHC velocity and fracture toughness are considered as probabilistic variables. In all the analyses, failure probabilities are calculated using the Monte Carlo (MC) simulation. It is clearly demonstrated from these analyses that failure probabilities are somewhat sensitive in size change of the pressure tube and the hydride precipitation temperature.

Derivation of Input Distributions for Probabilistic Evaluation Using In-Reactor Measurements

2012 ◽  
Author(s):  
Eric Nadeau
Keyword(s):  
Risk Reduction ◽  
Management Strategy ◽  
Reactor Core ◽  
Pressure Tube ◽  
Channel Management ◽  
Maintenance Strategy ◽  
Acceptance Criteria ◽  
Fuel Channel ◽  
Probabilistic Evaluation ◽  
Commercial Operation

Candu Energy Inc. (former commercial operation of AECL) has developed probabilistic tools to support nuclear plant operators with a risk-based fuel channel management strategy. One such tool is used to evaluate the probability of pressure tube rupture resulting from pressure tube to calandria tube contact and hydride blisters. This tool assumes that PT rupture occurs when delayed hydride cracking (DHC) initiates in a blister. The objectives of the probabilistic assessments are to: • Determine the overall risk of PT rupture in the reactor core for comparison with the acceptance criteria. • Determine the risk of PT rupture for specific fuel channels to assist in the development of an inspection/maintenance strategy. • Evaluate the risk reduction that would result from different fuel channels inspection/maintenance scenarios. • Optimize inspection/maintenance programs. The distributions of the most critical input distributions can be derived by benchmarking against in-reactor measurements. Two benchmark methods were developed to take advantage of the recent advancements in the accuracy of the inspection tool that measures the gap profile between the PT and the CT.

Experiences in Examining and Dispositioning Flaws in Zr-2.5Nb Pressure Retaining CANDU Reactor Components

2002 ◽  
Author(s):  
Andrew Celovsky ◽  
John Slade
Keyword(s):  
Service Life ◽  
Reactor Core ◽  
Pressure Tube ◽  
Candu Reactors ◽  
Periodic Inspection ◽  
Pressure Tubes ◽  
Surveillance Programs ◽  
Non Destructive ◽  
Limited Service ◽  
Non Destructive Techniques

CANDU reactors use Zr-2.5 Nb alloy pressure tubes, as the primary pressure boundary within the reactor core. These components are subject to periodic inspection and material surveillance programs. Occasionally, the inspection program uncovers a flaw, whereupon the flaw is assessed as to whether it compromises the integrity of the pressure-retaining component. In 1998, such a flaw was observed in one pressure tube of a reactor. Non-destructive techniques and analysis were used to form a basis to disposition the flaw, and the component was fit for a limited service life. This component was eventually removed from service, whereupon the destructive examinations were used to validate the disposition assumptions used. Such a process of validation provides credibility to the disposition process. This paper reviews the original flaw and its subsequent destructive evaluation.

A Probabilistic Integrity Assessment of Flaw in Zirconium Alloy Pressure Tube Considering Delayed Hydride Cracking

2003 ◽  
Vol 17 (08n09) ◽  
pp. 1587-1593 ◽  
Author(s):  
Sang Log Kwak ◽  
Joon Seong Lee ◽  
Young Jin Kim ◽  
Youn Won Park
Keyword(s):  
Probabilistic Analysis ◽  
Nuclear Reactor ◽  
Failure Criteria ◽  
Pressure Tube ◽  
Initial Surface ◽  
Tube Material ◽  
Integrity Assessment ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Hydride Cracking

In the CANDU nuclear reactor, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the nuclear fuel bundles and heavy water coolant. Pressure tubes are major component of nuclear reactor, but only selected samples are periodically examined due to numerous numbers of tubes. Pressure tube material gradually pick up deuterium, as such are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC), which is the characteristic of pressure tube integrity evaluation. If cracks are not detected, such a cracking mechanism could lead to unstable rupture of the pressure tube. Up to this time, integrity evaluations are performed using conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. This paper describes failure criteria for probabilistic analysis and fracture mechanics analyses of the pressure tubes in consideration of DHC. Major input parameters such as initial hydrogen concentration, the depth and aspect ratio of an initial surface crack, DHC velocity and fracture toughness are considered as probabilistic variables. Failure assessment diagram of pressure tube material is proposed and applied in the probabilistic analysis. In all the analyses, failure probabilities are calculated using the Monte Carlo simulation. As a result of analysis, conservatism of deterministic failure criteria is showed.

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.

Quantitative Characterization of Hydride Spacing for Cohesive-Zone Modelling of Fracture Toughness in Zr-2.5Nb Pressure Tubes

2016 ◽  
Author(s):  
Cheng Liu ◽  
Leonid Gutkin ◽  
Douglas Scarth
Keyword(s):  
Fracture Toughness ◽  
Cohesive Zone ◽  
Cohesive Zone Model ◽  
Model Development ◽  
Threshold Value ◽  
Pressure Tube ◽  
Zone Model ◽  
Empirical Distributions ◽  
Hydride Formation ◽  
Pressure Tubes

Zr-2.5Nb pressure tubes in CANDU 1 reactors are susceptible to hydride formation when the solubility of hydrogen in the pressure tube material is exceeded. As temperature decreases, the propensity to hydride formation increases due to the decreasing solubility of hydrogen in the Zr-2.5Nb matrix. Experiments have shown that the presence of hydrides is associated with reduction in the fracture toughness of Zr-2.5Nb pressure tubes below normal operating temperatures. Cohesive-zone approach has recently been used to address this effect. Using this approach, the reduction in fracture toughness due to hydrides was modeled by a decrease in the cohesive-zone restraining stress caused by the hydride fracture and subsequent failure of matrix ligaments between the fractured hydrides. As part of the cohesive-zone model development, the ligament thickness, as represented by the radial spacing between adjacent fractured circumferential hydrides, was characterized quantitatively. Optical micrographs were prepared from post-tested fracture toughness specimens, and quantitative metallography was performed to characterize the hydride morphology in the radial-circumferential plane of the pressure tube. In the material with a relatively low fraction of radial hydrides, further analysis was performed to characterize the radial spacing between adjacent fractured circumferential hydrides. The discrete empirical distributions were established and parameterized using continuous probability density functions. The resultant parametric distributions of radial hydride spacing were then used to infer the proportion of matrix ligaments, whose thickness would not exceed the threshold value for low-energy failure. This paper describes the methodology used in this assessment and discusses its results.

Technical Basis of a Code Case to Provide a Strain-Based Acceptance Limits for Service Level D Evaluation of Piping Systems Under Section III of ASME Boiler and Pressure Vessel Code

2019 ◽  
Author(s):  
Jie Wen ◽  
Timothy M. Adams
Keyword(s):  
Dynamic Load ◽  
Pressure Vessel ◽  
Service Level ◽  
Historical Background ◽  
Division I ◽  
Test Results ◽  
Acceptance Criteria ◽  
Piping Systems ◽  
Acceptance Limits ◽  
Technical Basis

Abstract Section III, non-mandatory Appendix F and mandatory Appendix XXVII provide the evaluation rules for Service Level D loads. Appendix F, article F-1341.2 permits plastic analysis, but provides only a stress-based acceptance criteria. Strain-based acceptance criteria for use in Appendix XXVII is under development. Strain-based acceptance criteria for Level D service limits for piping would be beneficial, especially, for regions where extremely high seismic events exist. A Code Case for strain-based acceptance criteria for level D service limits of piping is being developed in Section III of the ASME Boiler and Pressure Vessel Code. This paper presents the technical basis of the strain-based acceptance criteria Code Case. The authors reviewed extensive literatures which were published over last 50 years to build the foundation of the Code Case. A historical background of the seismic rules is given to explain why the load cycles are limited to 20 as in Section III, Division I, Subsection NB/NC/ND-3622.2 This paper also explains the consideration of reversing dynamic load and non-reversing dynamic load and how they are combined in the analysis. Comparison to current Appendix F (Appendix XXVII) acceptance criteria and some existing test results are provided to show the acceptability and conservatism of this Code Case.

Technical Basis for Revised Flaw Acceptance Criteria Under IWB-3610 of ASME Section XI

2005 ◽  
Author(s):  
Russell C. Cipolla ◽  
Keith R. Wichman
Keyword(s):  
Fracture Toughness ◽  
Lower Bound ◽  
Input Parameter ◽  
Fracture Initiation ◽  
Reference Curve ◽  
Acceptance Criteria ◽  
Class 1 ◽  
Reference Curves ◽  
Static Fracture ◽  
Technical Basis

This paper describes the change from the KIa to KIc in performing flaw evaluation for Class 1 ferritic steel components according to IWB-3610 and Appendix A of the ASME Boiler and Pressure Vessel Code, Section XI. The primary reason for making this change is to reduce the excess conservatism in the current flaw evaluation acceptance criteria. The Appendix A calculation methods can be used in accepting flaws detected as part of the plant inservice inspection program. The KIa and KIc reference curves represent two lower bound fracture toughness curves available in ASME Section XI. The KIa reference curve is a lower bound on all static, dynamic and arrest fracture toughness, whereas the KIc reference curve is a lower bound on static fracture toughness only. A similar change has already been implemented in the calculation of the heatup and cooldown pressure-temperature (P-T) limit curves, which are also based on fracture mechanics analysis. The P-T limits are developed according to Appendix G of Section XI, which employ similar methods to those of Appendix A. A key input parameter to the P-T calculations is the lower bound fracture toughness curve, KIc (prior to change in Appendix G, the lower bound curve KIR was used, which was equivalent to KIa). Based on the work described in this technical basis paper, Section XI recently approved the change from KIa to KIc in IWB-3610. In addition, changes to the use of KIc for fracture initiation, changes were also made to IWB-3613, which provides acceptance criteria for flanges and other shell regions near structural discontinuities. These changes clarified the scope of the article as to what discontinuities are covered and a redefinition for the minimum temperature from RTNDT + 60°F (RTNDT + 33°C) to just RTNDT. These changes are also discussed in this paper. The changes will appear in the 2005 addenda to the 2004 Code Edition for Section XI.

Solution of a Sample Problem Related to Revision 1 of Code Case N-830

2017 ◽  
Author(s):  
Mark Kirk ◽  
Steven Xu ◽  
Cheng Lui ◽  
Marjorie Erickson ◽  
Yil Kim ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Master Curve ◽  
Nuclear Reactor ◽  
Sample Problem ◽  
Asme Code ◽  
American Society ◽  
Technical Basis ◽  
Direct Use ◽  
Current Code

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case (CC) N-830. CC N-830 permits the direct use of fracture toughness in flaw evaluations as an alternative to the indirect/correlative approaches (RTNDT-based) traditionally used in the ASME Code. The current version of N-830 estimates allowable fracture toughness values in the transition regime as the 5th percentile Master Curve (MC) indexed to the transition temperature T0. The proposed CC N-830 revision expands on this capability by incorporating a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J–R) used currently, or useful in, a flaw evaluation for conditions ranging from the lower shelf through the upper shelf. Papers presented in previous ASME Pressure Vessel and Piping (PVP) Conferences since 2014 provide the technical basis for these various toughness models. This paper contributes to this overall CC N-830 documentation suite by presenting the results of a sample problem run to assess the proposed revision of the CC. The objective of the sample problem was (1) to determine if the revised CC was written with adequate clarity to permit different engineers to accurately and consistently calculate the various allowable toughness values described by the equations in the CC, (2) to assess how these allowable toughness values would be used to calculate allowable flaw depths using standard ASME SC-XI approaches, and (3) to compare allowable flaw depths calculated using established Code practices (RTNDT-based) to those calculated using proposed CC practices (T0-based). The sample problem demonstrated that (1) the CC was written with sufficient clarity to allow different engineers to arrive at the same estimated value of allowable toughness, (2) the latitude associated with the provisions of the ASME Code pertinent to estimation of allowable flaw depth are responsible for some differences in the allowable flaw depth values reported by different participants, and (3) current Code estimates of allowable flaw depth are far more conservative (that is: smaller) than values estimated by the candidate CC methods based on the MC, this mostly due to the generally-conservative bias of the Code’s RTNDT & KIc approach. The candidate CC methods provide much more consistent conservatism than current Code approaches for all conditions in the operating nuclear reactor fleet via their use of an index temperature (T0) defined by actual fracture toughness data and a temperature dependence defined by those data. The WGFE is continuing to evaluate candidate approaches to estimate allowable toughness values for CC N-830 using a T0-indexed Master Curve. Associated work is addressed by two companion papers presented at this conference.

Sign in / Sign up

Export Citation Format

Share Document