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.

The CANDU 1 (CANada Deuterium Uranium) reactor core consists of 380–480 horizontal Zr-Nb pressure tubes, which support fuel bundles and provide pressurized heavy water cooling. The pressure tubes are supported by fuel channel annulus spacers, which maintain the gap between the hot pressure tube and colder calandria tube while providing a means of leak detection through the annulus gas system. Research and testing in this area have shown that spacer material degradation in later life operation could impact the ability of the component to meet its design requirements. This paper presents a fitness-for-service strategy that could be utilized in demonstrating continued safe operation of these components. Fitness-for-service is based on analysis of crush tests on ex-service spacers to determine the load carrying capacity projected into the future and endurance tests to determine fatigue life. This paper describes these technical approaches and their application in fitness-for-service evaluation of spacers in CANDU operating plants to satisfy requirements for an annulus spacer surveillance program under Clause 12.5 of the CSA Standard N285.4-14.

During a Loss of Coolant Accident (LOCA) in a Boiling Water Reactor (BWR), subcooled water flows past jet pump assemblies located in the annular region between the Reactor Pressure Vessel (RPV) and the shroud as it moves toward the break location and is subsequently discharged from the RPV. Flow loads caused by such an event are required design basis loads that must be considered for BWR internal components. A comparison is presented between BWR annulus flow loads calculated by the 2D potential flow methodology using complex variables and the 3D finite element method using the heat transfer analogy described in PVP2016-63091 [1]. The comparison demonstrates that greater solution fidelity is available using the FEA method. The results also demonstrate the importance of applying the appropriate velocity correction factor when flow loads are computed using complex variable techniques.

We consider a hanging cantilevered pipe conveying water within a water-filled container; the upper portion of the pipe is surrounded by a rigid cylindrical tube of larger diameter, forming an annular fluid-filled region around the pipe. Two flow configurations are investigated : (a) water enters the pipe at its clamped end and flows downwards, discharging at its free end into the container; the fluid exits the container by flowing upwards in the annulus and out; (b) the reverse flow arrangement: water enters the system at the upper end of the annulus and exits by flowing upwards in the pipe. The dynamics of the system is studied theoretically and experimentally for both configurations. The analytical models utilized are outlined and the experiments are described. Theory and experiment find that the system loses stability at sufficiently high flow velocity by flutter or static divergence.

Piping vibration had been observed in one of our refinery’s reactor inlet piping for several decades. Vibration levels in inlet piping for reactor ‘D’ and ‘E’ were highest, relative to those in reactor ‘A’, ‘B’, and ‘C’. To cope with the vibration, design changes to small-bore branch connections had been implemented to reduce susceptibility to the vibration. A recent increase in production demand made the vibration levels more evident and a production constraint was imposed after an MOV gas seal failure. Analysis identified the root-cause as flow-excited acoustic resonance of (almost) coaxial closed side branches in the flow path. The selected vibration mitigation solution involved installing a perforated annulus in the main line, in front of the mouth of the (almost) coaxial closed side branch acoustic resonator. Before fabricating and installing the perforated annulus, it was decided to evaluate its expected performance by means of computational fluid dynamics (CFD) and structural stress finite element analysis (FEA). This paper gives an account of the selection of the perforated annulus as the preferred vibration mitigation solution and its evaluation by means of high-performance computing CFD and FEA. The CFD and FEA analysis showed that the perforated annulus would perform as intended and mitigate the piping vibration. The perforated annulus was fabricated and installed in the inlet piping for reactor ‘D’. Piping vibration was observed to be mitigated, even when flowing above the design rate. The perforated annulus vibration mitigation solution was replicated in the inlet piping for reactor ‘E’. The production constraint has since been lifted.

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.

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.

During a Loss of Coolant Accident (LOCA) in a Boiling Water Reactor (BWR), subcooled water flows past jet pump assemblies located in the annular region between the Reactor Pressure Vessel (RPV) and the shroud as it moves toward the break location and is subsequently discharged from the RPV. Flow loads caused by such an event are required design basis loads that must be considered for BWR internal components. In previous works [1, 2], the three dimensional (3-D) flow field problem was simplified to be a 2-D problem by assuming the radial velocity variations to be negligible. The 2-D problem was solved using complex function techniques by assuming a potential flow. Further, the velocity field had to be suitably scaled up to account for the presence of components such as the jet pumps in the annulus. In order to solve the problem in the realistic environment of a populated annulus, this paper illustrates a methodology where Finite Element Analysis (FEA) is used to perform a 3-D potential fluid flow calculation utilizing the analogy that exists between steady state heat transfer and potential flow problems. For an ideal fluid, the potential flow and irrotational flow assumptions will result in the Laplace equation as the governing equation for the velocity field. This is the same equation that governs the steady state heat transfer in any domain of interest where the temperature field is determined by solving the Laplace equation and applying the appropriate boundary conditions. Once the analogy between steady state heat transfer problems and potential flow problems governed by Laplace equation can be established, any commercially available finite-element code may be employed to solve such fluid flow problems involving complicated regions of interest by employing elements meant to solve heat transfer problems. For illustration purposes, a LOCA flow problem will be solved using Finite Element Model (FEM) thermal elements and compared against 2-D flow field results.

This paper discusses the design of a hardened enclosure for high pressure extreme temperature (HPET) pneumatic testing of seals. The seal test fixture is located within a blast-fragment barrier (BFB) in case of a failure of the test fixture. The BFB is located within a hardened test room since the BFB has a 1-in annulus air gap to allow for cabling access to the test fixture, through which shock and gas pressure can propagate out into the room. A 30.5in 3 (0.51) volume charged with 15,000psi (1034bar) of compressed nitrogen under a temperature range of −50°F to 400°F was used to produce the worst-case energy release in the event of a catastrophic failure of the seal fixture. In order to assess the structural capacities of the BFB and test room, an equivalent axisymmetric computational fluid dynamics (CFD) model of the test fixture, BFB, and test room was used to estimate the fragment velocity and applied blast load time histories to the BFB and test room. The BFB was assessed using non-linear dynamic finite element analysis. All of the materials are modeled using a plasticity-based piecewise linear hardening constitutive model, with rate dependency and softening, fit to each material yield, tensile, and elongation limit. The seal fixture generated fragment interactions with the interior surfaces of the BFB model using slideline and pinball contact algorithms.

The performance of fiber-reinforced polymer composites as thermal insulation for marine riser structures was investigated. The objective was to insulate steel piping contained in the riser pipe annulus to reduce the heat transfer to adjacent systems within the riser structure. Modeling was conducted to explore the effectiveness of the composite insulation. In addition, a scaled-down experimental setup was developed and used to validate the modeling results. In this contribution the modeling work, the fabrication of the insulation and the experiments are described.

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

In the disease syringomyelia, fluid-filled cavities, called syrinxes, form in the spinal cord. The expansion of these pathological pressure vessels compresses the surrounding nerve fibers and blood supply, which is associated with neurological damage. We investigate the spinal wave-propagation characteristics, principally to serve as a reference for more anatomically-detailed models. The spinal cord is modeled as an elastic cylinder, which becomes an annulus containing inviscid fluid when a syrinx is included. This is surrounded by an annulus of inviscid fluid, representing the cerebrospinal fluid occupying the subarachnoid space, with an outer rigid boundary approximating the dura mater. The axisymmetric harmonic motion is solved as an eigenvalue problem. We present dispersion diagrams and describe the physical mechanism of each wave mode. We identify potentially damaging syrinx fluid motions and tissue stress concentrations from the eigenvectors. Finally, we determine the dependence of each wave mode on syrinx radius and cord tissue compressibility.

This paper investigates the turbulence-induced vibration of a circular beam in annular pipe flow. Vibrations induced by turbulence are one of the causes of fatigue and fretting wear in process environments. Although the small-scale vibrations are normally not leading to immediate failure of structural components, they typically result in long term damage. To predict the amplitude of these subcritical vibrations, current methods require an accurate description of the incident pressure field. However, measurements of cross-spectral pressure fields in annular geometries are rare. Models to describe the pressure field have a tendency to provide only descriptive information, after a series of experiments have been performed. Therefore this paper aims to predict the pressure field numerically, by means of wall-resolved large-eddy simulations. In order to validate this approach the flow field of an experiment available in literature is computed. In the conditions simulated, water is flowing at 10 m/s in an annulus with a hydraulic diameter of 1.27cm. Pressure correlations obtained from the computations are compared to descriptive models such as the Corcos and Chase models. The numerical power spectra are also compared to experimental spectra.

In the seismic analysis practice, the calculation of modal response has traditionally been limited to a cutoff frequency of about 33 Hz based on United States Nuclear Regulatory Commission (US NRC) Regulatory Guide (RG) 1.60 [1] response spectra. The structural response in higher modes is calculated as a missing mass correction by static analysis. Seismic ground motions at several sites (such as Central and Eastern United States) exhibit high frequency content, up to about 100 Hz. Additionally, the reactor building vibratory (RBV) loads that result from the suppression pool hydrodynamic loads due to loss of coolant accident (LOCA), and the annulus pressurization (AP) load from a postulated pipe break at the reactor pressure vessel (RPV) safe ends and shield wall generate peaks at frequencies in excess of 100 Hz. The qualification of safety equipment supported in the reactor building needs to reflect these high frequency motions. Extracting frequencies and mode shapes up to zero period acceleration (ZPA) frequencies in these cases may not be practical or economical. Therefore, the cutoff frequency criteria for these types of high frequency loads need to be evaluated so that the analysis produces a representative and a reasonably conservative response. In this study, the equipment response is described in terms of stress quantities, member forces, and moments resulting from the solution up to a cutoff frequency. The responses are compared to the full solution up to the ZPA frequency under hydrodynamic and AP loads using the Response Spectrum Method. The cutoff frequency is deemed adequate if the ratio of the truncated response considering missing mass to the full response is 90% or greater. The internal strain energy (or its surrogate kinetic energy) for all modes with frequencies below the cutoff is also studied to assess the missing strain energy in modes in excess of the cutoff. The evaluation presented also examines how well the strain energy correlates with calculated stresses.

During the refueling and maintenance outage in August 2011 at Leibstadt Nuclear Power Plant in Switzerland, the inspection of the hydrostatic bearings of the two identical recirculation pumps revealed a deep circumferential erosion groove on the inside surface of each of the bearing journals. The bearing journals are made of austenitic stainless steel. The cylindrical journal is welded to the back shroud of the impeller and surrounds the internal stationary heat exchanger of the pump by forming a narrow fluid filled annulus. The location of material removal was the same as in the year 2004 when similar wear damage was fixed by build-up welding. The plant decided to repair the damage during the subsequent outage in 2012. However, the Swiss Federal Nuclear Safety Inspectorate in return required the plant to identify the precise erosion mechanism, to ensure the structural integrity of the journals by taking into account the rate of material removal from 2004 up to the 2012 outage, and to include provisions for the early detection of a journal failure. This paper summarizes the previous as well as the latest results of different inspections, investigations, evaluations, and analyses done to meet the requirements of the Swiss regulatory authority. The results show that, from a safety-related and an operational availability perspective, it is acceptable not to repair the damaged bearing journals prior to the 2012 outage.

An assessment was completed to address the failure of internal thermal sleeve weld for reheat condensate nozzle of steam generators. The plant is operated by Ontario Power Generation (OPG) in Pickering, and has CANDU ® 6 type reactors. The objective of assessment was to evaluate the effect of the failed weld on the overall structural integrity of the nozzle for the defined operating service conditions. The fitness for service of the steam generator nozzle was demonstrated by comparing the maximum stress ranges of the initial nozzle design with the failed weld nozzle configuration under the same service conditions. Two nozzle configurations were considered for this assessment. One configuration represents the original shape with no leakage at weld indicating as-designed condition. Transient heat transfer and the stress analyses were performed according to the defined service limits. Another configuration completed for the faulty condition in which weld is failed and thermal sleeve separated. The same transients as the first configuration were applied, but the leakage was introduced at the thermal sleeve weld. The effect of leakage was considered by changing the convection heat transfer coefficient in annulus area between the external side of sleeve and internal surface of the nozzle. Critical locations on the nozzle were identified for the whole transient cycles, and assigned different stress lines. The maximum and minimum stress intensity ranges of the initial nozzle design and the cracked weld nozzle design were compared for these stress lines. It was concluded that the thermal sleeve weld failure with the conservatively postulated leakage flow provides better results in terms of stress ranges compared to the as-design condition. The thermal shield was over constrained in as-design condition. And for the fitness for service evaluation in was decided to leave the failed weld in-service without repairing it.

Acoustic loads caused by a Recirculation Line Break Loss of Coolant Accident are a required design basis event that must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps. This event causes acoustic and fluid loads on BWR internals. These loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during in-service inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. These models generally do not enable accurate assessment of the variation of acoustic loading on jet pumps away from the break plane. Previous methods conservatively apply the acoustic loading predicted for the jet pump nearest the break for all jet pumps. Insight can be gained and lower loading may be justifiable, for jet pumps away from the break, using methods that enable accurate acoustic load predictions for all jet pumps in the BWR annulus. This paper presents the results of Recirculation Line Break acoustic loads analyses of a typical BWR and investigates the variation of acoustic loading between all jet pumps in the annulus. The paper also presents the results of preliminary sensitivity studies performed to identify which geometric characteristics of the BWR annulus have the most significant effect on the resulting acoustic load predictions. The analyses documented in this paper are performed using acoustic finite element analysis.

An amine tower was inspected and was shown to have wall loss over the majority of the circumference. A repair plan was developed which included welding onto the tower while the tower remained in operation. The two step repair plan first required welding two Inconel 625 rings to the SA516-70N steel tower and then welding a stainless steel sleeve to the Inconel 625 rings encapsulating the corroded area. Since the repair welds could be exposed to an amine environment if the steel tower corroded through, API Recommended Practice 945 (API RP 945) was used to aid in the qualification of the welding procedure. API RP 945 recommends a post-weld heat treatment (PWHT) to reduce hardness and relieve stress, but since the planned repair was to be made in-service, PWHT was not preferred. To address the hardness aspect a temper bead technique was used to successfully qualify a welding procedure, without PWHT, in accordance with 2006 NBIC and 2004 ASME Boiler and Pressure Vessel Code Section IX with hardness values below 200 Brinell. The temper bead welding procedure used Inconel 182 SMAW electrodes and required strict welding heat input control and weld bead placement. The heat input could be monitored by controlling the welding parameters or by using the run-out ratio diagram. The temper bead passes needed to be deposited in such a manner that the weld toe of the temper bead was no more than 3/32 in. (2.4 mm) away from the weld toe of the initial layer. To address the residual stress aspect the procedure qualification weld was thermo-mechanically modeled to predict the residual stress distribution on the inside surface of the amine tower. The repair procedure was performed on the operating vessel to emplace the straps and the sleeve. Shortly afterwards, the tower wall was breached and the internal environment reached the annulus inside the sleeve. Operation continued for several months until the replacement vessel was available. Once the vessel was removed from service, a section of the repaired area was examined for residual stresses and hardness in the carbon steel. The peak residual stresses were lower than predicted by the analysis from the qualification stage. However, the measured heat affected zone (HAZ) hardness was well above the desired level of 200 Brinnell. Analysis showed that the increased hardness level correlated with improper temper bead placement [i.e., temper bead to weld toe spacing greater than 3/32 in. (2.4 mm)] along with other indications of deviations from the qualified procedure.

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from the spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10–15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions. However, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

Jet pumps in a boiling water reactor (BWR) are located in the annulus region between the core shroud and the reactor vessel wall and provide core flow to control reactor power. Between 16 and 24 jet pumps are included in BWR/3 through BWR/6 plants, depending on the plant rating. The inlet mixer assembly of the jet pump is secured in place with a hold down mechanism called a jet pump beam. This beam is fabricated of alloy X-750 and tensioned to 58–74% of the yield stress of the material, depending on the beam design. In recent years, more attention has been placed upon inter-granular stress corrosion cracking (IGSCC) of alloy X-750 BWR internal components as a result of in-service cracking and failures. BWR plant owners have implemented actions to manage IGSCC of jet pump beams and assemblies through increased inspections and changes to process specifications for X-750. However, a thorough understanding of the flaw tolerance of the jet pump beam was not available to guide the periodicity of inspections as well as to define critical flaw sizes needed to validate the capability of inspection techniques. This paper describes a linear elastic fracture mechanics (LEFM) evaluation in which the flaw tolerance of the existing jet pump beam designs is established and used to recommend inspection frequencies for the jet pump beam. Industry operating experience is used to assess the credibility of the results obtained from this evaluation. This work illustrates an example of the use of LEFM to develop a technically defensible basis for the required inspection regions and the frequency of inspection for an alloy X-750 BWR internal component and helps to establish the necessary sensitivity of non-destructive examination technology to be used to examine the component.