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.

Application of “Leak Before Break” Assessment for Pressure Tube in Delayed Hydride Cracking

2010 ◽  
Vol 5 (4) ◽  
pp. 378-384
Author(s):  
Gintautas Dundulis ◽  
◽  
Albertas Grybėnas ◽  
Vidas Makarevicius ◽  
Remigijus Janulionis ◽  
...  
Keyword(s):  
Reactor Core ◽  
Zirconium Alloys ◽  
Pressure Tube ◽  
Cell Assembly ◽  
Deterministic Analysis ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Leak Before Break ◽  
Resistance To Brittle Fracture ◽  
Hydride Cracking

The Ignalina NPP uses an RBMK-1500 reactor, which is graphite-moderated with a water-cooled reactor core. The fuel cell assembly in the center of the moderator column consists of a pressure tube containing the fuel element assembly and through which coolant flows. Pressure tubes are made of Zr-2.5Nb zirconium alloys. Hydrogen absorbed by the zirconium alloy during corrosion is one of the factors determining pressure tube lifetime. If the pressure tube hydrogen concentration exceeds solubility limitations, delayed hydride cracking (DHC) may occur. Hydrides forming under certain conditions may reduce resistance to brittle fracture. Here we evaluate hydride influence on pressure tube fracture and the application of leak before break (LBB) for tubes with DHC. Deterministic analysis employing LBB concept used experimental data. Deterministic LBB analysis confirms that pressure tubes comply with LBB requirements.

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.

Leak-before-break in the pressure tubes of CANDU reactors

1990 ◽  
Vol 43 (1-3) ◽  
pp. 1-21 ◽  
Author(s):  
G.D. Moan ◽  
C.E. Coleman ◽  
E.G. Price ◽  
D.K. Rodgers ◽  
S. Sagat
Keyword(s):  
Candu Reactors ◽  
Pressure Tubes ◽  
Leak Before Break

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.

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.

Improvement of the Evaluation Accuracy for the Diameter Prediction of the CANDU Pressure Tubes

2020 ◽  
Author(s):  
Jong Yeob Jung
Keyword(s):  
Measured Data ◽  
Operating Conditions ◽  
Tube Diameter ◽  
Pressure Tube ◽  
Dramatic Improvement ◽  
Irradiation Creep ◽  
Candu Reactors ◽  
Irradiation Growth ◽  
Pressure Tubes ◽  
Dry Out

Abstract Measurements of the inside diameter of the pressure tubes in CANDU reactors have shown that the diameter has been increasing over time, and this phenomena has been explained as a creep phenomenon which is a kind of aging process of the pressure tube owing to the operating conditions of irradiation by neutron flux, high pressure, and high temperature over the plant life. The diameter expansion of the pressure tube has been regarded as a principle aging mechanism governing the heat transfer and hydraulic degradation within the primary heat transport system of the CANDU reactor. Diametrical expansion results in a reduction of the fuel cooling owing to the increased bypass flow, which increases the possibility of a fuel dry-out and thus limits the operating power of the reactor. In order to explain the mechanism of the creep phenomena of the pressure tube, traditionally the creep deformation has been modeled as a combination of thermal creep, irradiation creep and irradiation growth. However, this modeling approach is too complex to determine all parameters and constants which are relevant to the equation. In this research, we proposed a very simple approach for modeling the pressure tube diameter deformation in which the pressure tube diameter was modeled based on the measured data, flux distribution of each fuel channel and temperature variation inside the pressure tube. New rules were derived to determine the effect of flux and temperature distribution on the diameter expansion based on the measured data of pressure tube diameter. Results from applying the methodology show a dramatic improvement of the prediction accuracy of pressure tube diameter compared to the previous modeling results.

Pilot Study for Uncertainty Analysis in Probabilistic Fitness-for-Service Evaluations of Zr-2.5Nb Pressure Tubes: Identification of Influential Variables

2018 ◽  
Author(s):  
Christopher Manu ◽  
Suresh Datla ◽  
Leonid Gutkin
Keyword(s):  
Sensitivity Analysis ◽  
Pilot Study ◽  
Uncertainty Analysis ◽  
Pressure Tube ◽  
Output Variable ◽  
Probabilistic Evaluation ◽  
Pressure Tubes ◽  
Relative Change ◽  
Leak Before Break

Canadian Nuclear Standard CSA N285.8, “Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU® reactors”(1), permits the use of probabilistic methods when performing assessments of the reactor core. A non-mandatory annex has been proposed for inclusion in the CSA Standard N285.8, to provide guidelines for performing uncertainty analysis in probabilistic fitness-for-service evaluations within the scope of this Standard, such as the probabilistic evaluation of leak-before-break. The proposed annex outlines the general approach to uncertainty analysis as being comprised of the following major activities: identification of influential variables, characterization of uncertainties in influential variables, and subsequent propagation of these uncertainties through the evaluation framework or code. The application of the proposed guidelines for uncertainty analysis was exercised by performing a pilot study for one of the evaluations within the scope of the CSA Standard N285.8, the probabilistic evaluation of leak-before-break based on a postulated through-wall crack. The pilot study was performed for a representative CANDU reactor unit using the recently developed computer code P-LBB that complies with requirements of Canadian Nuclear Standard N286.7 for quality assurance of analytical, scientific, and design computer programs for nuclear power plants. This paper discusses the approach used and the results obtained in the first stage of this pilot study, the identification of influential variables. The proposed annex considers three approaches for identifying influential variables, which may be used separately or in combination: analysis of probabilistic evaluation outputs, sensitivity analysis and expert judgment. In this pilot study, local sensitivity analysis was used to identify and rank the influential variables. For each input variable in the probabilistic evaluation of leak-before-break, the local sensitivity coefficient was determined as the relative change in the output variable associated with a relative change of a small magnitude in the input variable. Each input variable was also varied across a large range to assess the linearity of the relationship between the input variable and the output variable. All relevant input variables were ranked according to the absolute value of their sensitivity coefficients to identify the influential variables. On the basis of the results obtained, the pressure tube wall thickness was found to be the most influential variable in the probabilistic evaluation of leak-before-break based on a postulated through-wall crack, followed by the fracture toughness of Zr-2.5Nb pressure tube material and the pressure tube inner diameter. The results obtained at this stage were then used at the second stage of this pilot study, the uncertainty characterization of influential variables, as discussed in the companion paper PVP2018-85011.

Exploratory Analysis to Estimate Axial Fracture Toughness for Zr-2.5Nb Pressure Tubes Using Test Data From Small Curved Compact Specimens

2017 ◽  
Author(s):  
Steven X. Xu ◽  
Kim Wallin
Keyword(s):  
Fracture Toughness ◽  
Prediction Model ◽  
Test Data ◽  
Late Life ◽  
Exploratory Analysis ◽  
Pressure Tube ◽  
Life Conditions ◽  
Pressure Tubes ◽  
Axial Fracture ◽  
Leak Before Break

Zr-2.5Nb pressure tubes are in-core, primary coolant containment of CANDU(1) nuclear reactors. Technical requirements for in-service evaluation of pressure tubes are provided in the Canadian Standards Associate (CSA) N285.8. These requirements include the evaluation of service conditions for protection against fracture of operating pressure tubes and demonstration of leak-before-break. Axial fracture toughness for pressure tubes is a key input in the evaluation of fracture protection and leak-before-break. The 2015 Edition of CSA N285.8 provides a pressure tube axial fracture toughness prediction model that is applicable to pressure tubes late life conditions. The fracture toughness prediction model in CSA N285.8-15 is based on rising pressure burst tests performed on pressure tube sections with axial cracks under simulated pressure tube late life conditions. Due to the associated high cost of testing and high consumption of pressure tube material, it is not practical to perform a large number of fracture toughness burst tests. On the other hand, more fracture toughness data is required to improve the existing pressure tube axial fracture toughness prediction model. There is strong motivation to estimate pressure tube axial fracture toughness using test data from small specimens. The estimated pressure tube fracture toughness using test data from small specimens can fill the gaps in the burst test toughness data, as well as provide information on material variability and data scatter. Against this background, an exploratory analysis of estimating pressure tube axial fracture toughness using test data from small curved compact specimens has been performed and is described in this paper. The estimated values of pressure tube axial fracture toughness using the test data from small curved compact specimens are compared with the measured toughness from burst tests of pressure tube sections with axial cracks to check the feasibility of this approach.

244Cm contributions to the alpha source term of CANDU reactors

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 106-115
Author(s):  
F. R. Greening
Keyword(s):  
Neutron Flux ◽  
Steam Generator ◽  
Power Law ◽  
Source Term ◽  
Pressure Tube ◽  
Candu Reactors ◽  
Activation Products ◽  
Pressure Tubes

Abstract In this report the expected rate of buildup of 244Cm in a CANDU neutron flux is evaluated and used to explain cases of high 244Cm in alpha-active samples from Bruce and Pickering Units. It is demonstrated that 244Cm is enriched on the surface of irradiated pressure tubes where it is associated with Zr/Nb activation products. It is further shown, using 94Nb as a fluence monitor, that the 244Cm/(239Pu + 240Pu) ratio for Bruce and Pickering irradiated pressure tube deposits exhibits a power law = 0.0042 (Fluence)3.1982. For non-pressure tube samples, such as feeder pipes, steam generator deposits and PHTS cruds, it is observed that Zr/Nb activation products are also associated with elevated 244Cm activities. Thus, based on the data presented in this report, the inference is that all CANDU Units may be expected to exhibit significant levels of 244Cm activity on PHTS surfaces, both in and out-of-core, with 244Cm/(239Pu + 240Pu) ratios significantly greater than one.

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