Improved Engineering Process-Zone Model for Evaluation of Structural Strength of Annulus Spacers in CANDU Nuclear Reactors

2017 ◽  
Author(s):  
Cheng Liu ◽  
Leonid Gutkin ◽  
Douglas Scarth
Keyword(s):  
Constitutive Relation ◽  
Structural Strength ◽  
Nuclear Reactors ◽  
Process Zone ◽  
Operating Time ◽  
Pressure Tube ◽  
Irradiation Damage ◽  
Zone Model ◽  
Engineering Process ◽  
Fuel Channel

The core of a CANDU®(1) pressurized heavy water reactor consists of a lattice of either 390 or 480 horizontal Zr-Nb pressure tubes, depending on the reactor design, which contain the nuclear fuel. Each pressure tube is surrounded by a Zircaloy calandria tube that operates at a significantly lower temperature. Fuel channel annulus spacers maintain the annular gap between the pressure tube and the calandria tube throughout the reactor operating life. To meet this design requirement, the annulus spacers must have adequate structural strength to carry the interaction loads between the pressure tube and the calandria tube. Crush tests performed on specimens from Inconel X-750 spacers, both non-irradiated and ex-service, have demonstrated that their structural strength had degraded with operating time due to irradiation damage. An engineering process-zone model was developed and used to analyze the spacer crush test results, and to predict the maximum load carrying capacities of the Inconel X-750 spacer coils, as described in the companion paper “Engineering Process-Zone Model for Evaluation of Structural Strength of Fuel Channel Annulus Spacers in CANDU Nuclear Reactors” presented at the PVP2017 Conference. The developed model is based on the strip-yield approach of a process zone with a uniform restraining stress that represents the fracture region surrounded by elastic material. This baseline process-zone model has been improved by allowing the restraining stress to evolve with the variation in the opening displacement in accordance with a traction-separation constitutive relation. The development of this improved engineering process-zone model incorporating a non-trivial traction-separation constitutive relation is described in this paper.

Engineering Process-Zone Model for Evaluation of Structural Strength of Fuel Channel Annulus Spacers in CANDU Nuclear Reactors

2017 ◽  
Author(s):  
Douglas Scarth ◽  
Steven Xu ◽  
Cheng Liu
Keyword(s):  
Structural Strength ◽  
Process Zone ◽  
Maximum Load ◽  
Pressure Tube ◽  
Irradiation Damage ◽  
Zone Model ◽  
Engineering Process ◽  
Fuel Channel ◽  
Load Carrying ◽  
Pressure Tubes

The core of a CANDU(1) (CANada Deuterium Uranium) pressurized heavy water reactor consists of a lattice of either 390 or 480 horizontal Zr-Nb pressure tubes, depending on the reactor design. These pressure tubes contain the fuel bundles. Each pressure tube is surrounded by a Zircaloy calandria tube that operates at a significantly lower temperature. Fuel channel annulus spacers maintain the annular gap between the pressure tube and calandria tube throughout the operating life. To meet this design requirement, annulus spacers must have adequate structural strength to carry the interaction loads imposed between the pressure tube and calandria tube. Crush tests that have been performed on specimens from as-received and ex-service Inconel X-750 alloy spacers have demonstrated that the structural strength of Inconel X-750 spacers has degraded with operating time due to irradiation damage. There was a need for an engineering model to predict the future maximum load carrying capacity of the spacer coils for use in Fitness-for-Service evaluations of spacer structural integrity. An engineering process-zone model has been developed and used to analyze the spacer crush test results, and provide predictions of the Inconel X-750 spacer coil future maximum load carrying capacities. The engineering process-zone model is described in this paper. The process-zone model is based on the strip-yield approach of a process zone with a uniform restraining stress representing the fracture region that is surrounded by elastic material.

Finite Element Verification of Engineering Equations for Prediction of Structural Strength of Annulus Spacers in CANDU Nuclear Reactors

2019 ◽  
Author(s):  
Preeti Doddihal ◽  
Dennis Kawa ◽  
Douglas Scarth ◽  
Yu Chen
Keyword(s):  
Finite Element ◽  
Heavy Water ◽  
Structural Strength ◽  
Process Zone ◽  
Bubble Formation ◽  
Operating Time ◽  
Pressure Tube ◽  
Operating Life ◽  
Engineering Approach ◽  
Pass Through

Abstract The core of a CANDU (CANada Deuterium Uranium) pressurized heavy water reactor includes several hundred horizontal fuel channels that pass through a calandria vessel containing the heavy water moderator. In each fuel channel, annulus spacers are used to maintain the gap between the cold calandria tube and the hot pressure tube, a pressurized vessel containing the nuclear fuel in contact with heavy water coolant. In order to carry the loads between the pressure tube and calandria tube, the annulus spacers are required to possess adequate structural strength throughout the operating life of the reactor. The Inconel X-750 spacers used in some reactor units are susceptible to irradiation induced degradation. As irradiation fluence increases with operating time, material embrittlement has been observed due to helium bubble formation in the X-750 spacer material. An engineering approach for assessing the structural strength of CANDU annulus spacers has been recently developed. When the ductility of the material is relatively low, the region susceptible to fracture under applied tensile stress may be adequately idealized as a strip-yield process zone surrounded by elastic material and associated with restraining stress. The engineering approach is based on applying the strip-yield process zone methodology to fracture at a nominally smooth surface. Finite element modeling was undertaken to simulate the strip-yield based fracture process zone. The finite element analyses and results are presented in this paper. The finite element results verify the engineering equations developed to assess the structural strength of annulus spacers.

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.

Development of Weight Functions for Modelling Delayed Hydride Cracking Initiation at a Blunt Flaw

2004 ◽  
Author(s):  
Douglas A. Scarth ◽  
Joanna Wu ◽  
Ted Smith ◽  
Dennis M. Kawa
Keyword(s):  
Stress Distribution ◽  
Process Zone ◽  
Mouth Opening ◽  
Crack Mouth Opening Displacement ◽  
Weight Functions ◽  
Zone Model ◽  
Engineering Process ◽  
Opening Displacement ◽  
Wide Range ◽  
Cubic Polynomial

Delayed Hydride Cracking (DHC) in Zr-2.5 Nb alloy material is of interest to the CANDU (Canada Deuterium Uranium) industry in the context of the potential to initiate DHC at a blunt flaw in a CANDU reactor pressure tube. The material is susceptible to DHC when there is diffusion of hydrogen atoms to the flaw, precipitation of hydride platelets, and development of a hydrided region at the flaw tip. The hydrided region can then fracture to the extent that a crack forms, and is able to grow by the DHC crack growth mechanism. An engineering process-zone model for evaluation of DHC initiation at a blunt flaw that takes into account flaw geometry has been developed. The model is based on representing the stress relaxation due to hydride formation, and crack initiation, by an infinitesimally thin process zone. Application of the engineering process-zone model requires calculation of the stress intensity factor, and the crack-mouth opening displacement, for a fictitious crack at the tip of a blunt flaw. In the current model, calculation of these quantities is based on a cubic polynomial fit to represent the stress distribution ahead of the blunt flaw tip, where the stress distribution is generally calculated by finite element analysis. However, the cubic polynomial is not always an optimum fit to the stress distribution for very small root radius flaws, due to the large stress gradients near the flaw tip. Application of the weight function method will enable a more accurate representation of the flaw-tip stress distribution for the calculation of the stress intensity factor and the crack-mouth opening displacement. Weight functions for a crack at the tip of a blunt flaw in a thin wall cylinder have been developed for implementation into the engineering process-zone model. These weight functions are applicable to a wide range of blunt flaw depths and root radii, as well as a wide range of flaw-tip crack depths. The development and verification of the weight functions is described in this paper. The verification calculations are in reasonable agreement with alternate solutions, and have confirmed that the weight functions have reasonable accuracy for engineering applications of the process-zone methodology.

Comments on the Process Zone Methodology When Applied to Delayed Hydride Cracking in CANDU Zr-Nb Pressure Tube Material

2008 ◽  
Author(s):  
E. Smith
Keyword(s):  
Process Zone ◽  
Pressure Tube ◽  
Zone Model ◽  
Tube Material ◽  
Delayed Hydride Cracking ◽  
The Way ◽  
Hydride Cracking

The paper discusses the application of the process zone model to the problem of hydrided region formation and Delayed Hydride Cracking (DHC) in CANDU Zr-Nb pressure tube material. The special characteristics of the process zone approach, as used for the DHC problem, are highlighted, while making comparisons with the way in which it is more generally applied in other engineering situations.

Multi-length scale micromorphic process zone model

2009 ◽  
Vol 44 (3) ◽  
pp. 433-445 ◽  
Author(s):  
Franck Vernerey ◽  
Wing Kam Liu ◽  
Brian Moran ◽  
Gregory Olson
Keyword(s):  
Process Zone ◽  
Zone Model ◽  
Length Scale

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.

Stress Re-Distribution Induced by Creep Relaxation Around a Notch Loaded in Tension: Measurement and Modeling

2009 ◽  
Author(s):  
Brian W. Leitch ◽  
Nicolas Christodoulou ◽  
Ronald Rogge
Keyword(s):  
Heavy Water ◽  
Strain Field ◽  
High Stress ◽  
Natural Uranium ◽  
Thermal Creep ◽  
Three Dimensions ◽  
Pressure Tube ◽  
Primary Coolant ◽  
Fuel Channel ◽  
Creep Response

The majority of the pressure-retaining components in the core of a CANDU power generation system are manufactured from zirconium. The horizontal fuel channel components and the fuel bundles that contain the natural uranium fuel are manufactured using various grades of zirconium. The fuel channel consists of two concentric tubes; an internally pressurized tube (Zr-2.5%Nb) that contains the fuel bundles (Zr-4) and the re-circulating heavy-water primary coolant, enclosed by a larger diameter calandria tube (Zircaloy) that separates the pressure tube from the heavy-water moderator. Re-fuelling and other fuel management operations can create surface defects in the tubes and fuel bundle sheathing. Stress analyses of these small notches may indicate that, under certain conditions, cracks can be formed at the root of these notches. These flaws are locations of stress concentration in the internally pressurized tube and can initiate a failure mechanism known as Delayed Hydride Cracking. The anisotropic material properties of these zirconium components adds an additional level of complexity in an analysis. However, the occurrences of these life-limiting events appear to be minimized mainly due to beneficial contributors such as stress relaxation around the scratches. One of the most likely reasons for this relaxation is thermal creep. Previously [1], the measurement and modeling of thermal creep relaxation under constant displacement was examined using 2-D finite element (FE) models. This paper extends both the measurement and modeling of the relaxing stress/strain field to the more demanding boundary condition of constant applied load. Neutron diffraction is used to determine the changing strain field around a single notched, axially orientated specimen loaded in tension. This specimen orientation and loading configuration is modeled in three dimensions using a hybrid explicit FE program [2] that contains materials subroutines that describe high stress creep specially developed to simulate the highly anisotropic creep response of pressure tube materials. Despite the difficulty of obtaining precise delineation of the moving strain field, a good agreement between the measurements and the 3-D FE creep results is achieved. Using the creep subroutines, the FE models are used to examine the creep response of a single notched, transversely orientated specimen loaded in tension in the hoop direction.

Nonuniformity of crack-growth resistance and breakdown zone near the propagating tip of a shear crack in brittle rock: A model for earthquake nucleation to dynamic rupture

1990 ◽  
Vol 68 (9) ◽  
pp. 1071-1083 ◽  
Author(s):  
Mitiyasu Ohnaka
Keyword(s):  
Crack Growth ◽  
Constitutive Relation ◽  
High Frequency ◽  
Crack Growth Resistance ◽  
Earthquake Source ◽  
Source Process ◽  
Zone Model ◽  
The Earth ◽  
Shear Rupture ◽  
Propagating Shear

This paper reviews our recent studies on (i) slip failure nucleation, which leads to the mechanical instability that gives rise to a dynamically propagating shear rupture, (ii) constitutive behavior during the local breakdown process near the propagating tip of the slipping zone, and (iii) the physical modeling of the earthquake-source process based on the constitutive relation inferred from laboratory experiments. Laboratory studies were done using a simulated fault in rock in the brittle regime under a mode II crack-growth condition, to gain a deeper understanding of the earthquake-source process, which is considered to be dynamically propagating shear rupture in the earth. A stable but accelerating phase of nucleation locally precedes an unstable dynamically propagating rupture even in the brittle regime. The appearance of a sizable zone of such nucleation is related to a nonuniform distribution of the crack-growth resistance on the fault. The local shear strength degrades to a residual friction stress level with ongoing slip near the propagating tip of the slipping zone. This slip-dependent constitutive relation shows that there is a breakdown zone near the propagating tip over which shear stress, slip displacement, slip velocity, and slip acceleration are highly nonuniform. This nonuniformity is responsible for generating high-frequency elastic radiation. A model of the breakdown zone, which incorporates the laboratory-based constitutive relation, does not give rise to unrealistic singularities of slip acceleration and stresses at and near the dynamically propagating tip of the slipping zone. The breakdown zone model enables one to give a common interpretation to both small-scale slip failure in the laboratory and large-scale shear failure as earthquake faulting in the earth, and it can explain the earthquake-source strong motion characterized by the high-frequency content.

Steam-Reheat Option for SCWRs

2009 ◽  
Author(s):  
Eugene Saltanov ◽  
Romson Monichan ◽  
Elina Tchernyavskaya ◽  
Igor Pioro
Keyword(s):  
Thermal Efficiency ◽  
Power Plants ◽  
Nuclear Reactors ◽  
Low Cost ◽  
Thermal Power ◽  
Reaction Rates ◽  
Pressure Tube ◽  
Outlet Temperature ◽  
Chemical Production ◽  
Production Of Hydrogen

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30 – 35% to about 45 – 48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency a nuclear steam reheat has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure in experimental reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear-reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the nuclear steam reheat are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

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