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.

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.

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.

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 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.

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.

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.

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.

Deformation Behaviour of a Transversely Loaded Garter Spring

2004 ◽  
Author(s):  
Brian W. Leitch
Keyword(s):  
Finite Element ◽  
Heavy Water ◽  
Reactor Core ◽  
Pressure Tube ◽  
Fuel Channel ◽  
Helical Springs ◽  
Pressure Tubes ◽  
Non Linear ◽  
Rigid Component ◽  
Water Moderator

The CANDU power generation system is based on a natural uranium fuelled reactor with a heavy water moderator. A unique feature of the CANDU reactor is the horizontal fuel channel that allows on-line re-fuelling and fuel management. Pressure tubes containing the fuel bundles and pressurized heavy water coolant are the in-core component of the fuel channel assemblies. Calandria tubes span the length of the reactor core and provide passageways for the pressure tubes through the reactor core. The calandria and pressure tubes are each approximately 6 meters long. The calandria tube separates the heavy water moderator (∼80°C) from the pressure tube (∼300°C). Both tubes are subjected to gravity loads but the pressure tube carries the additional load of the fuel bundles as well as experiencing high temperature and irradiation induced material effects. The pressure tube deflects under the combined loading and areas of the pressure tube could come into contact with the calandria tube. This contact would limit the operating efficiency and lifetime of the fuel channel. To maintain a gap between the pressure and calandria tubes, helical springs manufactured from rectangular cross-section wire are placed over the pressure tube. These helical springs are known as garter springs and four such springs are spaced along the pressure tube. Initially, there is no contact between the springs and the calandria tube, but as gravity forces and creep effects begin to act, the pressure tube sags and garter spring/calandria tube contact occurs. As the pressure tube continues to deform, a portion of the pressure tube weight, fuel and coolant is transmitted through the garter spring onto the calandria tube. The calandria tube, in turn, begins to deflect under the applied stresses. This creep deformation of the fuel channel takes place over many thousands of operating hours. Eventually, creep induces a permanent vertical deformation (sag) in the fuel channel. The sag of a fuel channel is an important factor in the operation of the structure and many methods are used to determine the general response of the pressure tube/calandria tube/garter spring system. These methods assume the garter spring is a rigid component. This paper specifically examines the garter spring behaviour with respect to the non-linear material and contact response between the pressure tube/garter spring/calandria tube components. A three dimensional (3-D) finite element solid model of the garter spring is used to determine the non-linear response of the helical garter spring to the transverse forces applied from 3-D shell finite element models of the pressure and calandria tubes. Comparison with experimental, crushing tests on garter springs illustrate the analytical model is well behaved. Applying the operating load to the 3-D model shows that the garter spring’s transverse deformation is small and that assuming the garter spring is a rigid component is valid.

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