Integrated Analysis of Mechanical and Thermal Hydraulic Behavior of Graphite Stack in Channel-Type Reactors in Case of a Fuel Channel Rupture Accident

2002 ◽  
Author(s):  
Sergei L. Soloviev ◽  
Boris A. Gabaraev ◽  
Leonid M. Parafilo ◽  
Dmitry V. Kruchkov ◽  
Oleg Yu. Novoselsky ◽  
...  
Keyword(s):  
Computational Analysis ◽  
Pressure Tube ◽  
Integrated Analysis ◽  
The Core ◽  
Pressure Tubes ◽  
Confinement System ◽  
Integrated Code ◽  
Channel Type ◽  
Graphite Stack ◽  
Fluid Parameters

The paper discusses the methodology and a computational exercise analyzing the processes taking place in the graphite stack of an RBMK reactor in case of a pressure tube rupture caused by overheating. The methodology of the computational analysis is implemented in integrated code U_STACK which models thermal-hydraulic and mechanical processes in the stack with a varying geometry, coupled with the processes going on in the circulation loop and accident localization (confinement) system. Coolant parameters, cladding and pressure tube temperatures, pressure tube ballooning and rupture, coolant outflow are calculated for a given accident scenario. Fluid parameters, movement of graphite blocks and adjacent pressure tubes bending after the tube rupture are calculated for the whole volume of the core. Calculations also cover additional loads on adjacent fuel channels in the rupture zone, reactor shell, upper and lower plates. Impossibility of an induced pressure tube rupture is confirmed.

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.

scholarly journals A simplified analysis of the Chernobyl accident

2021 ◽  
Vol 7 ◽  
pp. 1
Author(s):  
Bertrand Mercier ◽  
Di Yang ◽  
Ziyue Zhuang ◽  
Jiajie Liang
Keyword(s):  
Numerical Models ◽  
List Type ◽  
Flash Evaporation ◽  
The Core ◽  
Large Size ◽  
Reaction Coefficient ◽  
Pressure Tubes ◽  
Void Effect ◽  
Reactivity Coefficient ◽  
Reactivity Increase

We show with simplified numerical models, that for the kind of RBMK operated in Chernobyl: The core was unstable due to its large size and to its weak power counter-reaction coefficient, so that the power of the reactor was not easy to control even with an automatic system. Xenon oscillations could easily be activated. When there was xenon poisoning in the upper half of the core, the safety rods were designed in such a way that, at least initially, they were increasing (and not decreasing) the core reactivity. This reactivity increase has been sufficient to lead to a very high pressure increase in a significant amount of liquid water in the fuel channels thus inducing a strong propagating shock wave leading to a failure of half the pressure tubes at their junction with the drum separators. The depressurization phase (flash evaporation) following this failure has produced, after one second, a significant decrease of the water density in half the pressure tubes and then a strong reactivity accident due to the positive void effect reactivity coefficient. We evaluate the fission energy released by the accident

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.

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.

Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation

2020 ◽  
Author(s):  
Eunhyun Ryu ◽  
Hangyu Joo ◽  
Seungyul Yoo ◽  
Jongyub Jung
Keyword(s):  
Dimensional Space ◽  
Three Dimensional ◽  
Safety Margin ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Main Concern ◽  
Thermal Hydraulics ◽  
Pressure Tubes ◽  
Fuel Pin ◽  
Heavy Water Reactor

Abstract Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.

Multi-Variable Engineering Model for Axial Delayed Hydride Cracking Growth Rate in Pre-Irradiated Zr-2.5%Nb

2009 ◽  
Author(s):  
Leonid Gutkin ◽  
Douglas A. Scarth
Keyword(s):  
Growth Rate ◽  
Regression Model ◽  
Test Temperature ◽  
Pressure Tube ◽  
Tube Length ◽  
Variable Model ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Absolute Test ◽  
Hydride Cracking

The growth rate of postulated delayed hydride cracks in CANDU Zr-2.5%Nb pressure tubes is an important material property required for flaw evaluations and leak-before-break assessments. It is monitored using surveillance pressure tubes according to the requirements of the Canadian Standards Association (CSA) Standard N285.4 [1]. Radial growth rate and axial growth rate are used to calculate the propagation of delayed hydride cracks in the through-wall direction and along the pressure tube length, respectively. The axial delayed hydride cracking growth rate had been previously found to increase exponentially with inverse absolute test temperature. This dependence had been described by an Arrhenius-type regression model with one explanatory variable. As more experimental results were obtained from surveillance pressure tubes, it has become possible to assess whether there may be statistically significant effects of other variables, which should be incorporated into the representative relation for the axial delayed hydride cracking growth rate. In this paper, multi-variable regression analysis has been used to develop an improved representative model for the axial delayed hydride cracking growth rate of irradiated Zr-2.5%Nb pressure tube material. The developed model explains approximately 93% of overall observed variation in the experimental data, and therefore has better predictive capabilities than the reference regression model with test temperature as a sole predictor. The developed multi-variable model is proposed to be incorporated into the scheduled revision (2010 edition) of the CSA Standard N285.8 as the representative predictive model.

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.

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.

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