Diametral Creep Research Articles

Investigating various factors affecting neutron spectrum in the pressure tube in pressurized heavy water reactor design

The fuel bundles in the pressurized heavy water reactor design (PHWR), such as the CANDU (Canada Deuterium Uranium), are placed inside horizontal fuel channels. Each fuel channel consists of concentric tubes called the pressure tube (PT) and the calandria tube (CT). Due to continuous exposure to extreme neutronics and thermal hydraulics conditions, the PTs undergo physical deformations in both the axial and radial directions, as the reactor ages. The deformations include diametral creep, wall thinning, sag, and axial elongation. One of the contributing factors to these deformations is the neutron spectrum in the PT, especially neutrons with energy greater than 1 MeV. In this paper, the sensitivity of the neutron spectrum in the PT to various operating conditions of a PHWR is evaluated. Sensitivity assessments have been completed on several parameters namely the coolant purity, coolant temperature, moderator purity, moderator temperature, and pressure tube diametral creep (PTDC). Since the deformations are inevitable, identifying parameters that have the greatest impact on the population of neutrons with energy greater than 1 MeV in the PTs would be important for fine-tuning the operational conditions which are conducive to reducing the rate of deformation. From this study, it is concluded that the coolant temperature and PTDC are the two variables having the most impact on the fast energy neutron population in the pressure tubes.

Preliminary assessments of variations in the form-function-based maximum pin power in CANDU analyses

In the current CANDU full-core physics analysis, the maximum pin power within a bundle is determined by multiplying the bundle power obtained from solving the three-dimensional neutron diffusion equation with a ratio of the highest pin power to the average pin power, obtained from burnup-dependent infinite lattice calculations. Since the geometry of the fuel channel changes during the life of the reactor, the ratio of the highest to the average pin power changes as well. The objective of the current paper is to quantify such variations. Two sources of variation are considered: (i) pressure tube diametral creep (PTDC); and (ii) the placement of the fuel bundle inside the fuel channel in the lattice model. The results from the study indicate that the maximum pin power could vary up to 1.2%, depending on the amount of the PTDC, the fuel burnup, and the fact that a concentric or non-concentric lattice model is used. Therefore, it is recommended that the lattice model used for deriving the form function be as consistent as possible to the core condition being analysed, especially with respect to the PTDC value.

Multilevel modeling of diametral creep in pressure tubes of Korean CANDU units

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

FINITE ELEMENT ANALYSIS OF AN EMPTY 37-ELEMENT CANDU® FUEL BUNDLE TO STUDY THE EFFECTS OF PRESSURE TUBE CREEP

CANDU fuel bundles experience plastic deformations over time, and the horizontal configuration of the bundle in a crept pressure tube (PT) causes coolant to bypass the sagged lower half of the bundle. Bundle segments where the flow is limited may become more susceptible to dryout due to reactor aging. A finite element model of a 37-element fuel bundle was constructed using the commercial finite element software ANSYS to study the mechanical deformation behaviour of the bundle to maintain a coolable geometry. The main focus was on the contact between the fuel elements and between the fuel elements and PT. The complexity of the model due to all the contact pairs necessitated the use of high-powered computing hardware. Contact was demonstrated between the appendages, and sensitivity of the deformation to different boundary conditions (BC) was investigated. In particular, the radial position where the elements were welded to the endplate significantly impacted the magnitude of the element bowing. Expanding the PT up to 8% diametral creep demonstrated the proper functioning of the spacer pads (SP) and bearing pads in preventing sheath-to-sheath contact at the midplane and sheath-to-PT contact. However, the quarter plane was deemed to be the critical region due to the lack of SPs preventing excessive element bowing. This work has successfully illustrated the deformation of a CANDU fuel bundle, with contact, and its similarity with the bow profiles when compared with post-irradiation examination results and bundle heat-up tests.

Fluid flow in a diametrally expanded CANDU fuel channel – Part 2: Computational study

The current work presents investigations of flow through a 37-element CANDU nuclear fuel bundle residing within a deformed pressure tube. This scenario simulates the long-term effects of aging, whereby the pressure tube may experience up to ̃6% diametral creep, resulting in flow bypass and the concomitant increase in fuel temperature due to local undercooling. This work examines in high spatial detail the three-dimensional, three-component fluid velocity field through the fuel channel. A companion paper is dedicated to the experimental component of this work, which is based on Magnetic Resonance Velocimetry. In the present paper, computational fluid dynamic simulations have been performed with hydra using an implicit large eddy simulation to predict turbulent flow behavior. Together, an improved understanding has been gained in quantifying flow bypass, the evolution of geometry-induced inter-subchannel mixing, and various turbulent effects, such as recirculation, swirl, and separated flow. These capabilities are not possible with conventional experimental techniques or thermal-hydraulic codes. The overall goal of the combined works is to continue developing experimental and computational capabilities for future investigations to support nuclear reactor performance and safety.

Fluid flow in a diametrally expanded CANDU fuel channel – Part 1: Experimental study

The overall aim of this study is to develop and validate experimental and computational capabilities to provide high-resolution isothermal fluid velocity data for nuclear reactor investigations. The current work presents full-field velocity data in a 1:1 replica of a 37-element CANDU nuclear fuel bundle using the Magnetic Resonance Velocimetry (MRV) measurement technique. The diameter of the pressure tube is 6% larger than at beginning-of-life to simulate the long-term aging process in an actual CANDU nuclear reactor due to diametral creep at end-of-life conditions. In Part 1, the MRV technique is presented with a focus on measurement errors and the estimation of these errors. The three-dimensional, three-component mean velocity field inside the fuel channel, including the subchannel flow between the fuel elements, is provided at a resolution of 0.78 × 0.78 × 0.78 mm3. The MRV data is validated through bulk flow rate calculations and Laser Doppler Velocimetry measurements conducted at selected locations inside the flow system. Overall, the deviations in the full-field MRV data are low except for local differences in regions with high convective acceleration and high velocity. The measurement accuracy can be improved by applying newly developed MRV sequences that are insensitive to these effects. In this specific experiment, it is found that up to 35% of the coolant bypasses the fuel bundle in a 6% diametrally expanded pressure tube containing a single bundle. The data is used as a reference for a companion paper (Part 2), which presents a computational fluid dynamics approach using an implicit large eddy simulation. In conclusion, the presented experimental methodology provides accurate full-field mean velocity data in complex channel geometries. More realistic flow experiments in replicas of entire fuel channels can be investigated in detail with relatively little effort and in a short time.

Correlation between pressure tube thickness and expected discharged burnup in CANDU: Lattice calculations

CANDU (Canada Deuterium Uranium) is a pressurized-heavy-water reactor design where the fuel bundles are oriented horizontally inside the fuel channels. Each fuel channel is comprised of concentric tubes called the pressure tube and the calandria tube. As the reactor ages, due to continuous exposure to high neutron field, temperature, and pressure, a deviation from the concentric arrangement becomes more observable. As a result of irradiation and the presence of high pressure and temperature, the pressure tube undergoes changes such as diametral creep, sag, and axial elongation. The changes in the pressure tube geometries have an unfavourable impact on the operation of the reactor where the aged reactor overall power output must be reduced in order to maintain the same safety margin as the new reactor. One of the options to overcome this challenge is to utilize thicker pressure tubes. The impact of this change from the neutron economy point-of-view is evaluated in this paper; in particular its impact on the expected discharged burnup value. From this study, it is concluded that a percent increase in the thickness of the pressure tube leads to approximately one-half-of-a-percent decrease in the expected discharged burnup value.

Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Fracture Toughness Evaluation of Zr-2.5Nb Pressure Tubes Manufactured Employing Double Forging for PHWR700

The need for improvement of the in-reactor performance of Zr-2.5 % Nb pressure tubes arose because of the higher than expected diametral creep rates and variability in axial elongation observed in some of the pressure tubes of 220 MWe Indian pressurized heavy water reactors (IPHWR220). A systematic study revealed that the pressure tubes used in various IPHWR220 had variations in microstructure, mechanical properties, texture, and chemical composition. The forthcoming PHWR700 is expected to operate at a higher temperature than that of IPHWR220 pressure tubes coupled with ∼3 % partial boiling toward the outlet end. The IPHWR220 pressure tubes were manufactured by NFC (Hyderabad, India) using the combination of extrusion-double pilgering with intermediate annealing and with 20 % final cold work. PHWR700 pressure tubes have been manufactured using a combination of forging-extrusion-single pilgering routes. The fracture behavior of cold-worked and stress-relieved Zr-2.5Nb pressure tube material manufactured employing a forging route for PHWR700 is described in this work as a function of hydrogen content (4 and 60 wppm), temperature (25°C–300°C), and sample location. Fracture toughness tests were carried out as per ASTM E1820-06 [ASTM E1820-06: Standard Test Method for Measurement of Fracture Toughness—Designation, Annual Book of ASTM Standards, ASTM International, West Conshohocken, PA] using curved compact tension specimens of 17 mm width. The crack growth was along the axial direction of the tube and was measured using the direct current potential drop (DCPD) technique. The fracture toughness of as-received samples show weak dependency on test temperature, whereas samples having 50 ppm hydrogen show typical S curve behavior and regain toughness above 150°C. The variation in fracture toughness across tube length was within scatter band.

Artificial Neural Network Modelling of In-Reactor Diametral Creep of Zr2.5%Nb Pressure Tubes of Indian PHWRs

A model is developed to predict the in-reactor diametral creep in the Zr–2.5%Nb pressure tube of Indian Pressurized Heavy Water power reactors (PHWR) using Artificial Neural Network (ANN). The inputs of the neural network are alloy composition of the tube (concentration of Nb, O, N and Fe), mechanical properties (YS, UTS, %EL), temperature and fluence whereas diametral creep rate is the output. Measured diametral creep rate data from the sampled pressure tubes operating in Indian PHWRs at Rajasthan Atomic Power Station (RAPS 2), Kakrapar Atomic Power Station (KAPS 2) and Kaiga Generating Station (KGS) are employed to develop the model. A three-layer feed-forward ANN is trained with Levenberg–Marquardt training algorithm. It has been shown that the developed ANN model can efficiently and accurately predict the diametral creep of pressure tube. Results show the high significance of O concentration and mechanical properties in determining diametral creep rate.

Surface profiling with high density eddy current non-destructive examination data

The pressure tubes (PT) in CANDU (CANada Deuterium Uranium) reactors undergo creep induced deformation due to operating pressure, temperature and radiation conditions. While global deformation of the tube in the form of elongation and diametral creep is well characterized and monitored by station inspection systems, local PT deformation and the presence of inner surface artifacts due to wear are not as directly monitored, but can still provide additional information of fuel channel condition. A surface profiling technique for monitoring local deformation and identification of surface wear using an eddy current probe mounted in a small (50mm×25mm) planar probe body is presented. The sensitivity of the eddy current probe to small lift-off variations combined with high density C-Scan information is used to extract information on smoothly varying local deformation as well as monitor more significant wear on the inner surface of pressure tubes. Vector separation of components permits independent identification of axial and circumferential surface features. Analysis of this data can be used to characterize local PT deformation due to constrictions at fuel bundle ends and loaded garter spring spacers, as well as identify areas where shallow mechanical wear has occurred. Examples of the features that may be identified are presented.

Variations in nanomechanical properties of back-end Zr–2.5Nb pressure tube material

The Zr–2.5Nb pressure tube alloy used in Canada Deuterium Uranium (CANDU) nuclear reactors consists of a dual-phase microstructure produced by a multi-step manufacturing process. The metallurgical characteristics of the pressure tubes influence their in-reactor behavior, especially with respect to reactor life-limiting properties such as diametral creep. In order to predict the in-reactor behavior of pressure tubes, a greater understanding of the influence of tube-to-tube variations in metallurgical factors such as texture, grain size, and β-phase percentage would be greatly beneficial. In this paper, a testing method combining high temperature nanoindentation with subsequent electron channeling contrast imaging (ECCI) in the scanning electron microscope (SEM) is presented to compare two back end pressure tube off-cuts. Nanoindentation was used to measure local mechanical properties such as hardness, modulus, and strain rate sensitivity. Post-examination of residual indents with ECCI allowed for the correlation of microstructural characteristics to nanomechanical properties. A difference in hardness was observed between the two tubes in the axial normal plane, which was correlated to differences in β-phase area percentage and/or morphology.

PREDICTION OF DIAMETRAL CREEP FOR PRESSURE TUBES OF A PRESSURIZED HEAVY WATER REACTOR USING DATA BASED MODELING

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

DIAMETRAL CREEP PREDICTION OF THE PRESSURE TUBES IN CANDU REACTORS USING A BUNDLE POSITION-WISE LINEAR MODEL

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors

Abstract In an Advanced CANDU Reactor (ACR) (ACR is a registered trademark of Atomic Energy of Canada Limited), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure, and flux than those in a CANDU (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactor, and accordingly require a thicker wall (6.5 mm for ACR versus 4.2 mm for CANDU). In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a study to model and forecast the performance of these thicker pressure tubes has been undertaken. One of the main requirements for the pressure tube is to have low diametral creep. Based on previous experience with CANDU reactor pressure tube performance and manufacture, an assessment of the grain structure and texture of the ACR pressure tubes indicates that the in-reactor creep deformation will be improved. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor.

Predicting Diametral Creep of the Pressure Tubes in CANDU Reactors Using Fuzzy Neural Networks

Pressure tube (PT) creep is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS) in Canada deuterium uranium reactors. PT diametral creep affects the thermal hydraulic characteristics of coolant channels and the critical heat flux (CHF). CHF is a key parameter in determining the critical channel power, which is used in the trip setpoint calculations of regional overpower protection systems. This paper aims to predict PT diametral creep using the measured signals of the HTS by applying fuzzy neural networks (FNNs) according to operating conditions. The FNN model was optimized in terms of its fuzzy rules and parameters by a genetic algorithm combined with the least-squares method. Informative data that demonstrate the system's characteristic behavior were selected to train the FNN model using the subtractive clustering method. The proposed FNN model for predicting PT diametral creep was verified using the operating data of the Wolsong Unit 1 nuclear power plant in Korea. It was known that the FNN could predict the PT diametral creep accurately. Statistical and analytical uncertainty analysis methods were applied to the models and their uncertainties were evaluated using 60 sampled training and optimization data sets, as well as two fixed test data sets.

Effect of process variables on texture and creep of pressurised heavy water reactor pressure tubes

Zr-2.5 wt%Nb pressure tubes are used for coolant channels of Pressurised Heavy Water Reactors (PHWRs). These pressure tubes are lifetime components of the reactor and have to sustain extremely harsh conditions of temperature, pressure and neutron irradiation during service. One of the major life limiting factors for pressure tubes is diametral creep. This causes dilation of the pressure tube leading to flow bypass and inefficient heat removal from fuel bundles. This underscores the importance of producing pressure tubes with higher creep resistance. The primary metallurgical parameters controlling the creep strength of Zr-2.5Nb pressure tubes are texture, microstructure, grain size, dislocation density and alloying additions. This includes the strengthening of beta phase due to niobium enrichment. Heat treated Zr-2.5Nb pressure tubes have been reported to have the highest creep strength. This paper briefly discusses the effects of various processing parameters on the microstructure and texture of pressure tubes, which enhance their creep resistance.

Measurement and analysis of diametral deformation in irradiated Zr-2.5%Nb pressure tube

Internal diameter measurements have been carried out on pressure tube S-07 of KAPS II and fifteen RAPS II pressure tubes. These data have been analyzed to derive empirical correlation for predicting diameter change in pressure tube during irradiation. The correlations are derived for the diametral strain rates in terms of the stress, temperature and fluence for KAPS II and RAPS II pressure tubes which indicated essentially same functional dependences with a slight change in the prefactor (8.55 vs 12.35). This paper presents the details of the study carried out on modeling of diametral creep in the pressure tubes.

Mechanical Properties of Zr-2.5Nb Pressure Tubes Made from Electrolytic Powder

Abstract Pressure tubes in CANDUCANada Deuterium Uranium (CANDU) is a registered trademark of Atomic Energy of Canada Limited and RBMK reactors are fabricated from Zr-2.5Nb alloy. This paper describes the mechanical properties of tubes used in power reactors made by four routes using electrolytic powder as the base material. The microstructures developed by each route are distinct: CW: cold-worked material consists of flattened α-Zr grains surrounded by a skin of β-Zr phase; used in CANDU 6 reactors. CW-A: material that was cold-worked and annealed at 540°C contains elongated α-Zr grains mixed with equiaxed α-Zr grains and particles of β-Nb phase; used in all RBMK 1000 reactors. TMT-1: material quenched from the (α+β)-Zr phase into water follow by cold-working consists of α′-phase and between 10 and 20 % of untransformed α-phase; used in RBMK 1500, Ignalina 1. TMT-2: material quenched from the (α+β)-phase into argon-helium gas mixture followed by cold-working consists of Widmanstätten α-phase and untransformed α-phase. This material is used in RBMK 1500, Ignalina 2. The CW and TMT-2 tubes have a higher proportion of grains with basal plane normals in the transverse direction, FT of 0.52 to 0.57, than in the radial direction, FR of 0.38, while quenched and annealed materials (TMT-1 and CW-A) have similar values of FT and FR, about 0.38 in quenched materials and 0.41 in annealed materials. Transverse tensile strength, crack growth resistance, dJ/da, and axial crack velocity, VH, of delayed hydride cracking (DHC) were evaluated, using standard techniques, between 250 and 300°C on as-fabricated materials. In-reactor creep deformation was evaluated from measurements of tube diameter in RBMK 1000s, RBMK 1500s and two CANDU 6 power reactors. Strength and crack growth resistance were measured on TMT-1 and TMT-2 tubes removed from Ignalina NPP Units 1 and 2 after 12–17 years of in-reactor service. As-received cold-worked material had the highest strength; the annealed material had the lowest strength while the quenched materials had intermediate strength. Irradiation increased the strength by about 200 MPa in all four materials. Although DHC is sensitive to texture and the distribution of the β-Zr phase, the dominating factor controlling crack velocity appears to be material strength: with an increase of strength by a factor of about two, VH increased by a factor of 30. Since harmful trace elements were well controlled during manufacturing, other factors affecting crack growth resistance could be assessed. Again, strength appeared important; dJ/da declined approximately linearly with increase in strength induced by irradiation, decreasing from about 350 to 100 MPa as the strength increased from about 250 to 850 MPa. The exception was TMT-2 material where crack growth resistance was maintained after irradiation. TMT-2 material also had good diametral creep resistance in-reactor, attributed to both its texture and grain structure. The other three materials had similar creep resistance controlled mostly by their texture.

Effect of Texture on Anisotropic Creep of Pressurized Zr-2.5Nb Tubes

In this paper we report the stress dependence of the diametral and axial creep rates of cold-worked Zr-2.5Nb tubes with three different combinations of texture and microstructure in biaxial tests at 623K. One type of tube has a predominantly radial texture and it exhibits a high diametral creep rate and negative long term axial creep rate. A second type, with a predominantly transverse texture and a very fine elongated grain structure exhibits an intermediate diametral strain rate and a positive axial strain rate. The third type, with predominantly transverse texture and coarser elongated grain structure exhibits the lowest diametral creep and a positive axial strain rate. The stress exponents for the diametral creep are in the range 5-7, but those for axial creep vary from 0.7 to 7.3. The creep anisotropies were calculated with the polycrystalline model SELFPOLY-7, assuming that slip is the strain producing mechanism. The calculations give qualitative agreement with the experimental results.