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.

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.

Cycle-Wise Process-Zone Model for Prediction of Delayed Hydride Cracking Initiation Under Flaw-Tip Hydride Ratcheting Conditions

2018 ◽  
Author(s):  
Steven X. Xu ◽  
Jun Cui ◽  
Douglas A. Scarth ◽  
David Cho
Keyword(s):  
Structural Integrity ◽  
Process Zone ◽  
Operating Conditions ◽  
Zone Model ◽  
Evaluation Procedures ◽  
Fuel Bundle ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Technical Basis ◽  
Hydride Cracking

Flaws found during in-service inspection of Zr-2.5Nb pressure tubes in CANDU(1) reactors 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. A key structural integrity concern with in-service blunt flaws is their susceptibility to delayed hydride cracking (DHC) initiation, particularly for debris fretting flaws under flaw-tip hydride ratcheting conditions. Hydride ratcheting conditions refer to situations when flaw-tip hydrides do not completely dissolve at normal operating temperature, and accumulation of flaw-tip hydrides occurs with each reactor heat-up/cool-down cycle. A significant number of in-service flaws are expected to be under hydride ratcheting conditions at late life of pressure tubes. DHC initiation evaluation procedures based on process-zone methodology for flaws under hydride ratcheting conditions are provided in CSA (Canadian Standards Association) N285.8-15. The process-zone model in CSA N285.8-15 predicts whether DHC initiation occurs or not for given flaw geometry and operating conditions, regardless of the number of reactor heat-up and cool-down cycles. There has been recent new development. Specifically, a cycle-wise process-zone model has been developed as an extension to the process-zone model in CSA N285.8-15. The cycle-wise process-zone model is able to predict whether DHC initiation occurs or not during a specific reactor heat-up and cool-down cycle under applied load. The development of the cycle-wise process-zone model was driven by the need to include flaw-tip stress relaxation due to creep in evaluation of DHC initiation. The technical basis for the development of the cycle-wise process-zone model for prediction of DHC initiation under flaw-tip hydride ratcheting conditions is described in this paper.

A Statistical-Based Fatigue Crack Initiation Model for Axial Flaws in Zr-Nb Pressure Tubes

2014 ◽  
Author(s):  
Cheng Liu ◽  
Douglas Scarth ◽  
Alain Douchant
Keyword(s):  
Fatigue Crack ◽  
Crack Initiation ◽  
Rise Time ◽  
Structural Integrity ◽  
Fatigue Crack Initiation ◽  
Mechanical Damage ◽  
Fuel Bundle ◽  
Test Parameters ◽  
Pressure Tubes ◽  
Criteria For Evaluation

Flaws found during in-service inspection of CANDU Zr-2.5Nb pressure tubes include fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws, mechanical damage flaws and crevice corrosion marks. The CSA Standard N285.8 contains procedures and acceptance criteria for evaluation of the structural integrity of CANDU Zr-2.5Nb pressure tubes containing flaws. One of the requirements is to evaluate the flaws for fatigue crack initiation. There was a need to develop a statistical-based model of fatigue crack initiation at flaws for use in deterministic and probabilistic assessments of Zr-2.5Nb pressure tubes. A number of fatigue crack initiation experiments have been performed on notched specimens from irradiated and unirradiated Zr-2.5Nb pressure tube material with a range of hydrogen equivalent concentrations. These experiments were performed in an air environment and included temperature and load rise time as test parameters. The test data has been used to develop a statistical-based model of fatigue crack initiation at flaws that covers the effects of flaw root radius, load rise time and irradiation. This paper describes the development of the statistical-based model.

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.

Statistical Assessment of Crack Initiation at Simulated Flaws due to Hydrided Region Overload in CANDU Zr-2.5%Nb Pressure Tube Material

2011 ◽  
Author(s):  
Leonid Gutkin ◽  
Douglas A. Scarth
Keyword(s):  
Stress Concentration ◽  
Crack Initiation ◽  
Applied Stress ◽  
Pressure Tube ◽  
Experimental Program ◽  
Tube Material ◽  
Statistical Assessment ◽  
Hydride Formation ◽  
Pressure Tubes ◽  
Threshold Stress Intensity Factor

CANDU Zr-2.5%Nb pressure tubes are susceptible to formation of hydrided regions at the locations of stress concentration, such as in-service flaws. When the applied stress acting on a flaw with an existing hydrided region exceeds the stress at which the hydrided region has been formed, hydrided region overload may occur. Probabilistic methodology is being developed to evaluate in-service flaws in the pressure tubes for crack initiation due to hydrided region overload. Statistical assessment of relevant experimental data on the overload resistance of Zr-2.5%Nb has been performed as part of this development. The results of this assessment indicate that the critical nominal stress for crack initiation due to hydrided region overload increases with increasing the nominal applied stress during hydrided region formation, decreasing the stress concentration factor and increasing the threshold stress intensity factor for initiation of delayed hydride cracking. These findings are consistent with our fundamental understanding of hydrided region overload, as well as with the previous modeling work by E. Smith, as referenced in the paper. The overload resistance also appears to increase with the number of thermal cycles in the course of hydride formation. The results of this assessment have been used to develop a preliminary probabilistic model to predict the critical stress for crack initiation due to hydrided region overload under ratcheting hydride formation conditions, as well as a comprehensive experimental program to further investigate the overload behavior of CANDU pressure tube material.

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.

Effect of Hydrogen Isotope and Concentration on Delayed Hydride Crack Growth Rates in Zr-2.5Nb Pressure Tubes

2010 ◽  
Author(s):  
Gordon K. Shek ◽  
Harry Seahra
Keyword(s):  
Test Temperature ◽  
Growth Rates ◽  
Hydrogen Isotope ◽  
Hydrogen Diffusion ◽  
Critical Length ◽  
Potential Drop ◽  
Axial Direction ◽  
Peak Temperature ◽  
Pressure Tube ◽  
Pressure Tubes

CANDU Zr-2.5 Nb pressure tubes are susceptible to a cracking mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation and fracture at a crack tip. As defense-in-depth, when DHC is postulated to have initiated from a flaw, it is required to demonstrate that the crack can be detected by the leak monitoring system and the reactor safely shut down before the crack reaches the critical length for pressure tube rupture. DHC growth rates (DHCR) in the axial direction of the tube are required for such leak-before-break assessment. In this test program, the effect of hydrogen isotope and its concentration on DHCR in an unirradiated Zr-2.5 Nb pressure tube is studied. Pressure tube sections were hydrided or deuterided to different concentrations (nominal concentrations of 60, 100 and 190 ppm by weight). For the deuterided tube sections, they contained about 10 ppm of hydrogen from the manufacturing process. The DHC growth rate tests were performed on fatigue pre-cracked curved compact tension specimens, machined from the hydrided or deuterided tube sections, in several stepper-motor controlled load frames with cracking being monitored by direct current potential drop and acoustic emission techniques. DHCR at three test temperatures (270°C, 240°C and 200°C) were obtained from each specimen with the test temperatures approached from a peak temperature of 330°C. Some specimens were tested with a peak temperature of either 370°C or 300°C. The two main conclusions from the study are: (1) DHCR are affected by the hydrogen in solution at the test temperature and not by the amount of bulk hydrides present. The hydrogen in solution at a given test temperature depends on the hydrogen concentration of the specimen, as well as the thermal history (peak temperature in the initial thermal cycle and the test temperature) as a result of the hysteresis of Terminal Solid Solubility between hydride dissolution during heating and precipitation during cooling. (2) The DHC growth rates of the hydrided material are higher than those of the deuterided material because of the higher diffusion rate of hydrogen than deuterium. The isotope effect of hydrogen on DHC growth rates depends on the test temperature, with no apparent effect at 200°C and about 37% difference at 270°C which is slightly below the factor of √2 expected from the mass law of diffusion. The observed temperature dependence could be due to the presence of about 10 ppm hydrogen in the deuterided specimens, which dominates the DHC process at 200°C but insufficient to have a large effect at 270°C. The implication of the observed isotope and concentration effect of hydrogen on DHC growth rates on leak-fore-break assessment of flaws in pressure tubes is discussed.

Development and Application of Oxide Wedging Crack Growth Model for CANDU Zr-2.5Nb Pressure Tubes

2016 ◽  
Author(s):  
Douglas A. Scarth ◽  
Preeti Doddihal ◽  
Monique Ip
Keyword(s):  
Crack Growth ◽  
Growth Model ◽  
Growth Mechanism ◽  
Pressure Tube ◽  
Operating Pressure ◽  
Metallographic Examination ◽  
Reactor Coolant ◽  
Pressure Tubes ◽  
Crack Growth Mechanism ◽  
Crack Growth Model

Surface breaking and subsurface manufacturing flaws have been detected by in-service ultrasonic examination in a number of CANDU reactor Zr-2.5Nb pressure tubes. The manufacturing flaws are oriented in the circumferential direction in the Zr-2.5Nb pressure tube, and are axially aligned along the pressure tube. Metallographic examination of these flaws in an ex-service Zr-2.5Nb pressure tube revealed a series of parallel circumferentially oriented discontinuous features oriented at an angle of nominally 40 degrees relative to the axial direction in the pressure tube. Since the surface breaking flaws are exposed to the reactor coolant, it was considered prudent to evaluate potential growth of the flaws by an oxide wedging crack growth mechanism. Oxide wedging crack growth is a slow crack growth mechanism that can occur when zirconium oxide forms on the crack faces due to a corrosion reaction with the reactor coolant. An oxide wedging crack growth model was developed to predict crack growth rates and future flaw sizes as a part of the fitness-for-service evaluation of a Zr-2.5Nb pressure tube containing this type of manufacturing flaw. The model was then applied to predict crack growth from manufacturing flaws that were detected in an operating pressure tube, and the evaluation results were used as part of the justification for continued operation.

Validation of Weight Function Based Process-Zone Model for Evaluation of Delayed Hydride Cracking Initiation in Zr-2.5Nb Pressure Tubes

2016 ◽  
Author(s):  
Steven X. Xu ◽  
Dennis Kawa ◽  
Jun Cui ◽  
Heather Chaput
Keyword(s):  
Weight Function ◽  
Process Zone ◽  
Matrix Material ◽  
Pressure Tube ◽  
Zone Model ◽  
Hydrogen Atoms ◽  
Fuel Bundle ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Hydride Cracking

In-service flaws in cold-worked Zr-2.5 Nb pressure tubes in CANDU(1) reactors are susceptible to a phenomenon known as delayed hydride cracking (DHC). The material is susceptible to DHC when there is diffusion of hydrogen atoms to a service-induced flaw, precipitation of hydrides on appropriately oriented crystallographic planes in the zirconium alloy matrix material, and development of a hydrided region at the flaw tip. The hydrided region could then fracture to the extent that a crack forms and DHC is said to have initiated. Examples of in-service flaws are fuel bundle scratches, crevice corrosion marks, fuel bundle bearing pad fretting flaws, and debris fretting flaws. These flaws are volumetric in nature. Evaluation of DHC initiation from the flaw is a requirement of Canadian Standards Association (CSA) Standard N285.8. This paper describes the validation of the weight function based process-zone model for evaluation of pressure tube flaws for DHC initiation. Validation was performed by comparing the predicted threshold load levels for DHC initiation with the results from DHC initiation experiments on small notched specimens. The notches in the specimens simulate axial in-service flaws in the pressure tube. The validation was performed for both un-irradiated and pre-irradiated pressure tube material.

The Effect of Load Reduction on Crack Initiation Behavior of Hydrides From Flaws in Zr-2.5Nb Pressure Tube Material

2010 ◽  
Author(s):  
Jun Cui ◽  
Gordon K. Shek
Keyword(s):  
Crack Initiation ◽  
Hydrogen Diffusion ◽  
Load Condition ◽  
Peak Stress ◽  
Constant Load ◽  
Pressure Tube ◽  
Load Reduction ◽  
Operating Pressure ◽  
Notch Tip ◽  
Reactor Operating

Flaws in Zr-2.5Nb alloy pressure tubes in CANDU® nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC). DHC is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth of the hydrided region, and fracture of the hydrided region at the flaw tip. One scenario of crack initiation is that the flaw-tip hydrides are formed and cracked at the same stress under constant reactor operating pressure. This is known as crack initiation under constant-load condition. Another scenario of crack initiation is that the flaw-tip hydrides are formed at the operating pressure and then cracked during a transient over-pressure. This is known as crack initiation under overload condition as the hydrides are subjected to a stress higher than the hydride formation stress. In some CANDU reactors, a 20% reduction in pressure is implemented during reactor cool-down. This paper examines the effect of pressure reduction, and hence load reduction, on flaw-tip hydride morphology and crack initiation behavior under constant-load and overload conditions. Experiments were performed on specimens of an unirradiated Zr-2.5Nb pressure tube, with 57 wt. ppm hydrogen concentration. The specimens contained machined V-notches with a root radius of 0.015 mm to simulate service-induced debris fretting flaws. The results indicate that the 20% load reduction increases the threshold stresses for crack initiation under constant-load and overload conditions. Finite element stress analyses were performed to determine the notch-tip stress distribution under constant-load and 20% load-reduction conditions. The load reduction lowers the notch-tip peak stress and shifts its location away from the notch surface. This is consistent with the notch-tip hydride morphologies observed using optical and scanning electron microscopy.

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