Search

CANDU (CANada Deuterium Uranium) reactors have many unique design features that play important roles during a severe accident, however analysis of such features using Light Water Reactor (LWR) specific computer codes is challenging. Severe accidents in CANDU involve complex thermo-mechanical deformation phenomena which differ from the phenomena present during LWR accidents. For example, during complete station blackout scenarios with a failure of all emergency measures, the pressure tubes may balloon or sag into contact with the surrounding calandria tubes (CTs) establishing a thermal conduction pathway for heat rejection to the large moderator water volume. As the moderator liquid evaporates or boils its level decreases until fuel channels become uncovered in the calandria vessel. The uncovered channels heat up quickly and the entire fuel channel assembly (fuel, pressure tube and calandria tube) will sag and possibly disassemble. During the disassembly process some channel components may fall to the bottom of the calandria while others may form a suspended debris bed supported by channels which are still submerged in moderator liquid. These phenomena impact event-timing, accident progression, hydrogen production and fission product release. In this work several mechanistic channel deformation models have been developed and integrated into RELAP/SCDAPSIM/MOD3.6 to provide a coupled treatment of the deformation phase for such postulated accidents. MOD3.6 is a new version of the RELAP/SCDAPSIM code being developed to support the analysis of Pressurized Heavy Water Reactors (PHWRs) under severe accident conditions. In this paper, the code system is used to simulate a postulated station blackout accident for a generic 900 MW CANDU plant. To reduce the uncertainty in the modeling of core disassembly and to overcome the memory constraints of the code, the simulation is broken into two phases with the first phase (i.e., from initiating event to the channel failure and depressurization) simulated using a full-plant RELAP5 model providing relatively high spatial fidelity of the entire heat transport system, and the second phase (i.e. continued from the end of the first phase until calandria vessel dryout) using alternative nodalization focusing on the calandria vessel and fuel channel components. The paper assesses the entire accident progression up to the point of calandria vessel dryout and performs sensitivity analysis on model parameters to assess their relative importance. [ABSTRACT FROM AUTHOR]

In the event of a complete Station Blackout (SBO) of a CANDU reactor, the current safety practice is to initiate depressurization early in the transient once a lack of power has been declared. This requires operator actions to manually open the main steam safety valves (MSSVs) on the secondary side initiating a crash-cool procedure. The depressurization of the secondary side allows make-up water to be supplied to the steam generator (SG) secondary-side thereby extending natural-circulation driven heat removal from the fuel in the primary heat transport system. Once depressurized there are several additional water sources available to replenish the secondary-side inventory, and in the event that emergency power cannot be restored emergency mitigating equipment (EME) are available to provide alternative water make-up. The objective of this paper is to examine the processes and phenomena involved during and after crash-cooling and compare these results to cases where operator actions are not credited. Simulations are performed until such time as the secondary-side inventory is stabilized from alternative water sources or until it is depleted. A detailed RELAP5 model of a 900 MW CANDU plant has been created including the primary heat transport system (PHTS), the feed and bleed system, the steam generator secondary side, the moderator system, and the shield water cooling system. The 480 fuel channels were grouped into 20 channels by elevation and channel power. The models were benchmarked against the 1993 loss-of-flow event at Darlington NGS and agreed with the station data within the reported measurement uncertainty. Then the models were used to simulate the Station Blackout accidents with loss of class IV, class III, and emergency power supplies. Five different scenarios with/without crash-cool and with different water make-up options are modeled and key sensitivities determined. The results show that the depressurization of the secondary side may create a situation where continuous natural circulation breaks down and intermittent buoyancy induced flows (IBIF) takes place. The RELAP5 predicted IBIF phenomena are discussed, as well as the limitations of the current RELAP5 code. The main focus of this paper is on the early stage of the accidents, i.e. when adequate steam generator secondary side inventory exists and where damage to the main heat transport system can be precluded. The results demonstrate that EME actions to maintain SG inventory are effective and ensure fuel and fuel channel integrity. [ABSTRACT FROM AUTHOR]

A CANDU fuel string consists of 12 or 13 fuel bundles that laid horizontally inside a pressure tube. Under the large amount of drag load induced by the coolant flow the behaviour of individual bundles are coupled together through friction and contact. Presence of large number of contact and friction constraints causes the dynamical behaviour of many mechanical systems, like the vibration of CANDU fuel string, to be very complex. In this paper a numerical method based on linear complementarity problem (LCP) is developed to simulate vibrational behaviour of such systems. Then the presented method is employed and for the first time, natural frequencies of a CANDU fuel string are obtained numerically. Knowing the natural frequency of the string is very beneficial and can help to mitigate and decrease the damages, improve new fuel designs and develop new safety standards. All the solid components are discretized in space domain by the means of finite element method. Bozzak-Newmark integration scheme is employed to discretize the system equations of motion in the time-domain. With the computational power available today, frictional contact among fuel elements via spacer pads, between fuel elements and the pressure tube via bearing pads, and between neighbouring fuel bundles via endplates are modelled and the response of the string is obtained. Two different fuel string consist of 12 and 13 bundles are studied in this paper. FFT analyses are performed and natural frequencies of the systems are extracted. Results show great agreement with experimental values. The effect of boundary conditions in the last endplate of the downstream bundle on the natural frequencies is also investigated. [ABSTRACT FROM AUTHOR]

The instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. In the light water reactor technology where batch fuelling practice is employed, the snapshot of power distribution during the cycle could be easily obtained by monitoring and simulating the core from the beginning to the end of the fuel cycle. Such calculation is more challenging for CANDU reactors. Considering the fact that CANDU utilizes the on-power refuelling to compensate for the reduction in reactivity due to fuel burnup, in CANDU fuel management analyses, snapshots of power and burnup distributions can be obtained by simulating and tracking reactor operation over an extended period using various tools such as the *SIMULATE module of the reactor fuelling simulation program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly and use such snapshot to perform additional reactor-physics related analyses. Presently such patterns could be generated by a program called RANDIS which is implemented within the *INSTANTAN module. An alternative approach called DERMAGA ( D istribut E d R ando M A ge generator using G enetic A lgorithm) has been developed to derive the patterned-channel-age model where an optimization algorithm is utilized to find patterns which produce representative power distributions in the core. Within DERMAGA, the genetic algorithm technique has been utilized as the optimization engine to find a quasi -optimal patterned-channel-age. It has been demonstrated that DERMAGA is a viable alternative tool to produce representative snapshots of daily core configuration. In the present paper which is Part II of a two-part paper, the robustness of DERMAGA is tested against various different sizes of basic grid used for generating the random patterned-channel-age. [ABSTRACT FROM AUTHOR]

Nuclear energy is considered a key alternative to overcome the environmental issues caused by fossil fuels. It offers opportunities with an improved operating efficiency and safety for producing power, synthetic fuels, delivering process heat and for multigeneration applications. The high-temperature nuclear reactors, although possess great potential for integration with thermochemical water-splitting cycles for hydrogen production, are not yet commercially established. Current nuclear reactor designs providing heat at relatively low temperature can be utilized to produce hydrogen using thermochemical cycles if the temperature of their thermal heat is increased. In this paper, a hybrid chemical-mechanical heat pump system is proposed for upgrading the heat of the Enhanced CANDU (EC6) reactor design to the quality required for the copper-chlorine (Cu–Cl) hybrid thermochemical water splitting cycle operating at 550–600 °C. A modification to the heat pump is proposed to bring the heat to temperature higher than 650 °C with operating coefficient of performance estimated as 0.65. • Upgrade of waste heat from EC6 reactor is feasible using chemical heat pumps. • High-temperature process heat can be produced at over 550 °C and up to 800 °C. • The proposed heat upgrading option is more promising than conventional electric heating. [ABSTRACT FROM AUTHOR]

Cryogenic distillation (CD) of hydrogen in combination with Liquid Phase Catalytic Exchange (LPCE) or Combined Electrolytic Catalytic Exchange (CECE) process is used for tritium removal/recovery from tritiated water from fission rectors like CANDU and is the proposed technology for ITER/DEMO. Tritiated water is being obtained after long time operation of CANDU reactors, or in case of ITER mainly by the Detritiation System. The cryogenic system of the Experimental Pilot plant for tritium and deuterium separation from ICSI Rm. Valcea consists of four distillation columns and significant effort is required in various batch mode operations for achieving high tritium concentration. Some problems have been experienced with the fourth column of the cascade regarding the heat transient transfer during start-up. This paper intent is to present a CAE/FEA (Computer Aided Engineering/Finite Element Analysis) thermal simulation of the column during the transitory thermal regime in order to investigate the temperature distribution along the column and its connections to other equipment. The simulation is done on the as-built design of the column and takes in consideration the actual configuration. This work provides an important platform to understand the thermal phenomena during cool down of a cryogenic distillation column besides finding the failures in thermal insulation or design flaws. The results of the simulation can be used as lessons-learned for future design work of cryogenic systems like Tritium Removal Facility from Cernavoda Nuclear Power Plant or ITER/DEMO Isotope Separation System. [ABSTRACT FROM AUTHOR]

• Cernavoda Tritium Removal Facility as tritium supplier for fusion activities is in line with ITER plans for construction. • Tritium quantities produced by CTRF are lower than those required by fusion activities, but CTRF can be considered as an important contributor. • Position of CTRF in EU may be considered as an advantage, all requirements for tritium management (storage for transportation, transfer arrangements, safety) can be easily met. National R&D for Cryogenics and Isotopes Separation (ICSI) completed the conceptual design of the Cernavoda Tritium Removal Facility (CTRF) in 2015. CTRF is located at CNE 1 1 Centrala Nucleara Electrica - NPP Cernavoda and is sized to process 40 kg/h heavy water from 2 CANDU reactors, with a detritiation factor of 100, over 40 years design life. CTRF removes tritium using liquid phase catalytic exchange (LPCE) paired with cryogenic distillation (CD). The design of CTRF uses expertise from ICSI (Pilot Plant for Tritium and Deuterium Separation) and from Canada (Kinectrics), together with experience from Wolsung TRF project, research tritium laboratories and industry. The CTRF design uses the most advanced TRF technology, including recent safety requirements for a tritium industrial facility as specified by the Romanian Regulator. Construction of CTRF is planned to start in 2020 and the detritiation of heavy water from first reactor is scheduled to begin in 2026. After this date tritium stored will become available to be used for fusion research and industrial facilities. First phase of detritiation process is to reduce the moderator tritium content from 65 Ci/kg to 10 Ci/kg, second phase to maintain the tritium concentration up to 10 Ci/kg and third phase to use CTRF to reduce tritium content as low as possible before decommissioning of the site. This paper will present the technologies used, provide a prediction of tritium production, and considerations for possible use of CTRF for supplying He-3, as a by-product of the detritiation process. [ABSTRACT FROM AUTHOR]

• Development of a custom built non-contact 3D profiling system for pressure tube replicas. • Fiber based FD-OCT configuration was used to design and develop the system. • Inner surface replicas of pressure tube with flaws were obtained using replica tool. • 3D surface profile of replicas with various flaws were imaged by the system. This paper describes a 3D profiling system for surface profiling of replicas obtained from pressure tube (PT) of PHWR/CANDU type reactors. This system is based on Optical Coherence Tomography (OCT) technique. Fiber based Fourier domain OCT configuration was used to design and develop the system considering the potential application of imaging radioactive PT replica samples. The sample scanning assembly of the system has been custom designed considering the cylindrical nature of PT replicas. Replica of inner surface of pressure tube was obtained on a replica plate using a custom built replica tool. Three PT replicas of non-radioactive pressure tubes with various flaws were obtained in different conditions. These replicas were imaged in 3D for identification and quantification of the flaws. This 3D profiling system termed as 'PT replica imaging system' is a promising tool which can be useful in periodic inspection of pressure tubes for safe operation of nuclear reactor. [ABSTRACT FROM AUTHOR]

Abstract: The objective of this paper was to study the phase change of the debris formed on the CANDU 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys Fluent code. Two models using Large Eddy Simulation and Phase Change Effective Convectivity Model (PECM) predict the conjugate, radiative and convective heat transfer inside and from the corium pool. LES require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. From the safety point of view it is very important to maintain a heat flux through the wall below the critical heat flux assuring the integrity of the calandria vessel, thus the paper results include temperature profiles and heat fluxes through calandria wall. The results predicted by the PECM approach are comparable within 10% with those obtained by the computational fluid dynamics method, proving that the PECM can be used to perform further sensitivity studies. [Copyright &y& Elsevier]

Abstract: The performance of a newly developed three dimensional (3-D) reactor kinetic model for CANDU reactor is evaluated in this paper. The evaluation is carried out in a closed-loop environment with the help of an existing reactor regulating system (RRS). The 3-D model is obtained based on a modal synthesis method, which represents the neutron flux distribution inside the reactor in both time and space. Throughout this paper, a special attention has been paid to compare the performance of the developed 3-D model with that of traditional coupled point kinetic models. The dynamic behavior of the reactor model in a practical load-following mode has also been examined. The accuracy of the model has been validated against actual plant measurements under both transient and steady-state conditions. Through the analysis and the simulation studies, it has convincingly demonstrated that the developed 3-D model has significant advantages over traditional coupled point kinetic models in terms of the improved accuracy and the higher resolution in modeling the reactor internal dynamics. [Copyright &y& Elsevier]

Abstract: Using separated heavy water as moderator and supercritical water (SCW) as coolant introduces challenge for CANDU-SCWR to get a negative coolant void reactivity (CVR), due to which the moderator thickness of the fuel channel is optimized in this paper. When SCW flows through the core, there is a rapid variation in SCW density, which is directly related to the neutron spectrum and subsequently to the power distribution, so the 3D core neutronics/thermal-hydraulics coupling is needed to accurately evaluate the core coolant density and power distribution. In this paper, the neutronics calculation is computed with 3D fine mesh diffusion code while the thermal-hydraulic calculation is based on single channel model, they are coupled with each other automatically by a link code. Further, the in-core fuel management can be simulated by the link code to search the equilibrium cycle. Based on these calculation models, a CANDU-SCWR equilibrium core is designed with a thermal power of 2540MW, the core equivalent diameter is 4.30m and the active length is 5.94m. A 3-batch fuel management scheme with a cycle length of 350 EFPD is used. The numerical results show that a high average outlet coolant temperature of 625°C is achieved with a maximum cladding surface temperature less than 850°C. The maximum linear heat generation rate is 50.6kW/m, the average discharged burnup is 38.1GWd/tU, and the CVR is negative throughout the cycle. [Copyright &y& Elsevier]

Abstract: With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past (). The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100–1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides. [Copyright &y& Elsevier]

Abstract: In this paper, we propose a new method for the cell local flux reconstruction for CANDU® applications. The method is based on the modulation method where the core local mesh flux is reconstructed by modulating the form function with an interpolated cell mesh flux along with an appropriate normalization technique. The reconstruction method was adapted to the diffusion finite difference method. In this paper we present: (a) the flux reconstruction method, (b) the verification of this method using two benchmarking problems and (c) its validation using the ZED-2 data generated from the defueled channel experiments. [Copyright &y& Elsevier]

Abstract: In CANDU® reactor design, the regional overpower protection (ROP) systems protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. The increase in fuel power could be caused by a localized power peaking within the core (for example, as a result of a certain reactivity device configuration) or a general increase in the core power level during a slow-loss-of-regulation (SLOR) event. This overpower could lead to fuel sheath dryout. In the CANDU® 600MW (CANDU 6) design, there are two ROP systems in the core, one for each fast-acting shutdown system. Each ROP system includes a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal assemblies. A new methodology for designing the detector layout for the ROP system, called the DETPLASA algorithm, has been developed recently. This method utilizes the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The evaluation of the trip setpoint (TSP) corresponding to each detector layout configuration (i.e., each history in the SA algorithm) is performed probabilistically using the ROVER-F code. In this evaluation, there are uncertainties related to both the detector components (i.e., related to the margin-to-trip) and to the fuel channel components (i.e., related to the margin-to-dryout). In this paper, the importance of these uncertainties on the outcome of the detector layout optimization process is evaluated. Some parametric studies have been performed to quantify the effect of uncertainties on the resulting detector layout. Two types of investigations have been performed. First, a given detector layout will be used to explicitly determine the effect of changing the uncertainty values. In this study, 343 sets of uncertainty values are used to produce the corresponding TSP values. The variation in the TSP values is analyzed. Second, three sets of uncertainty values (a subset of uncertainties from the first study) are used in independent DETPLASA executions. The resulting detector layout configurations will be examined to observe the effect of these uncertainties on the final design. Results from these investigations are presented in this paper. [Copyright &y& Elsevier]

In Part I of this paper, we optimized the CANDU-6 adjuster rods for reactors burning thorium-based fuel. Here we modify the liquid zone controllers assuming optimized adjuster rods for the same advanced cycles. The objective is to maximize the average exit burnup for these fuel cycles while preserving the reactor ability to control the power distribution inside the core during normal operation conditions. These goals are achieved by implementing a liquid zone controller iterative response model based on explicit diffusion calculations to evaluate the control system capacity to maintain zone, channel and bundle powers below their safety limits during pre-determined perturbations sequences. Based on the evaluation of spatial control metrics for on-power refueling, boric acid or heavy water doping of the light water in the controllers is selected. A full core model is also used to assess their response during adjuster rods withdrawal in shim mode. The results demonstrate that our approach works well for thorium-based cycles, while achieving excellent fuel management performances. The use of D 2 O -doped liquid zone controllers for the natural uranium cycle can also increase its average exit burnup by 0.7% without detrimental consequences on reactor control. [ABSTRACT FROM AUTHOR]

Abstract: In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the long-term coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris. [Copyright &y& Elsevier]

Abstract: Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors worldwide and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under appropriate operating conditions. In planning refurbishment of nuclear plants stations, a key concern is the longevity of existing SGs up to the 60-year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and refines its estimation based on data specific to CANDU operating conditions. The paper presents a more advanced Bayesian probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of potential occurrences of degradation on reliability of SG tubing. [Copyright &y& Elsevier]

Abstract: The paper presents the results of sub-channel analysis of CANDU–SCWR based on a wide review of heat transfer correlations. According to comparison with experiment data at different heat flux, Bishop Correlation is selected in SUBCHAN code to analyze CANDU–SCWR fuel channel. By detailed calculation of 43 fuel rods fuel channel in CANDU–SCWR, the paper gets the conclusion that the mass flux redistribution and reduction of heat-transfer coefficient at supercritical condition caused by the steep change of coolant density will limit the power of fuel channel in CANDU–SCWR. [Copyright &y& Elsevier]

Significant concerns after Fukushima disaster were raised on the potential of radioactive release from Spent Fuel Pool accidents. In case of CANDU type reactors a set of factors, such as the low discharge burnup of the spent fuel and the pool configuration with trays and racks placed on the bottom of the pool, contributes to the reducing of this potential. In this paper a peculiar state is analysed in relation with the refurbishment process, when a full core is added into the spent fuel pool, in a shorter time then the normal refuelling discharge. A prolonged loss of cooling is postulated in order to boil and evaporate the water. The claddings of the fuel bundles accidently entered in the air will be progressively heated and then degraded. Fission products will be released into the atmosphere of the spent fuel pool and, possibly, into the environment through the existing junctions. The paper investigates the source term formation at the level of the environment considering a prolonged accident without any interventions. The simulation is structured into three phases: de-watering of the pool, fuel degradation, transport of the radioisotopes to the environment. In order to estimate the maximum potential of the accident, three scenarios for the discharge of the core were considered, all of them requiring an enlarged cooling capacity in comparison with the normal refuelling, aiming a reduction of the duration of the refurbishment process to reduce the economic penalties. [ABSTRACT FROM AUTHOR]

Abstract: This paper presents a thermal hydraulic analysis of the thorium-based advanced CANDU Reactor (TACR-1300) fuel channel. The paper first uses CATHENA, a system thermal hydraulics code for CANDU reactors, to determine the boundary conditions of the TACR-1300 fuel channel. The boundary conditions are then passed to ASSERT-PV, a subchannel thermal hydraulics code for the CANDU industry, to obtain detailed thermal hydraulic information on the TACR-1300 fuel channel. Through the subchannel analysis, it is found that the onset of dry-out power (ODP) for the TACR fresh fuel channel is much less than that of the ACR-700 because thorium fuel rods as well as uranium rods are not heated and there is bypass-flow effect in the associated subchannels. The paper compares three thorium loading modes, and determines the best one for TACR. [Copyright &y& Elsevier]

Even with current computing capabilities, detailed full core three-dimensional (3-D) transport calculations are still not practical. However, if we are satisfied with knowing only the average values of spatial flux distributions, the 3-D diffusion solution will constitute the final solution. On the other hand, in reactor design and safety analysis, direct information about the local flux distribution for the heterogeneous assemblies is required to assess the design and determine the safety margins. For this reason, after having solved the full-reactor-core problem, we have to look into the possibilities of recovering in a second step the information on local properties of single heterogeneous assemblies. In particular, the detector readings at detector locations are derived using these global homogenized parameters by applying appropriate numerical methods such as advanced interpolations. In this paper, we propose a method based on flux reconstruction to calculate the simulated detector readings in three-dimensions with high fidelity. Data from detector readings are very important in ensuring optimal reactor operations as well as in detecting any deviations from normal operations. Thus, calculating the detector readings with high fidelity will allow improvements to operating and safety margins. To validate this method, comparisons between detector reading simulation results and measurements from an operating CANDU reactor will be conducted and results will be presented. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Pressure tubes within CANDU reactors are subject to frequent ultrasonic non-destructive examination to identify and characterize any defects that pose a risk of initiating Delayed Hydride Cracking (DHC), a well-known problem that occurs in zirconium components that are subject to high mechanical and thermal stresses. The analysis of the ultrasonic data gathered from the pressure tubes is often a long and repetitive process on the critical path to the restart of the reactor. Motivations for developing an automated system include saving time on the critical path, minimising human subjectivity from the process and increasing repeatability of measurements. An automated system providing decision support to analysts also reduces the risk of operator fatigue by minimising time spent processing routine defects. This paper describes a novel system for the automated analysis of CANDU pressure tube inspection data. The system automates the entire initial examination process including the complex decision-making procedures by implementing an expert system with an associated rule-base. Results from this system are detailed, illustrating the location and characterisation of defects and key features within the pressure tube with a high degree of accuracy while increasing repeatability through the removal of subjectivity in measurements. [ABSTRACT FROM AUTHOR]

The application of thorium fuel in advanced CANDU (CANadian Deuterium Uranium) reactors is an economic solution to sustainable development of nuclear energy. In this paper, the thermal hydraulic design and transient analysis of Thorium-based Advanced CANDU Reactor (TACR) heat transport system are conducted using self-programmed code CANTHAC (CANDU Thermal Hydraulic Analysis Code). In the code CANTHAC, the single channel model and the point kinetics model are adopted to simulate the core thermal hydraulic and neutronic behaviors. A distributed parameter model is adopted in the simulation of integral economizer U-tube steam generator (IEUTSG). Multi-region non-equilibrium model is applied to simulate the pressurizer. The characteristics of heat transport pumps were simulated by four-quadrant analogy curves. The code is verified with ACR-700 steady-state analysis. The calculation results agreed well with ACR-700 design values and the relative errors are all within 2%. With CANTHAC steady-state module, the preliminary thermal hydraulic design of TACR is conducted. A 1000 MW double loop reactor with 520 fuel channels is designed. Single loop mass flow rate was designed to be 3202 kg/s. Under the design scheme, the maximum cladding temperature and the maximum temperature are 340.2 °C and 1369.9 °C, which satisfied the design requirements. The transient analysis of TACR heat transport system was performed with CANTHAC transient module. SG feed water temperature reduction accident, and complete loss of Class IV accident were conducted. In the SG feed water temperature reduction accident, a new balance of neutron and thermal hydraulic is reached and the core power increases by 5% at last. In the complete loss of Class IV accident, the maximum fuel temperature reaches peak value at 1172 °C and still has some safety margin. The ROH (Reactor Outlet Header) pressure does not exceed 105% of the design value, which meet the safety criterion demand. [ABSTRACT FROM AUTHOR]

The Wolsung Nuclear Power Plant in Korea is under IAEA safeguards. During IAEA physical inventory verifications, the Optical Fiber radiation Probe System has been used since 2007 as the official instrument for verification of spent fuel. Its low sensitivity requires excessive time and labor to perform the IAEA safeguard activities and increases the operator load for preparing a dangerous and time-consuming secondary inspection. The objective of this paper is to develop a superior new optical fiber-based scintillation detector to deal with such issues. To increase the sensitivity of the instrument, a new optical fiber-based scintillation detector was developed by combining a p-terphenyl scintillator with a new type of optical fiber. A field test was carried out in the CANDU Wolsung Unit 1 to characterize the performance of the new detector. The field test showed that the OFPS using the new optical fiber-based scintillation detector has at least 100-times higher sensitivity and can also measure even very weak signals not measurable by the existing detector. From the experimental results, we can conclude that the newly developed OFPS detector would increase the effectiveness and efficiency of nuclear safeguards for verification of spent fuel in wet storage pools of CANDU reactors. • A new optical fiber-based scintillation detector has been assessed for spent-fuel safeguards. • The detector is made of a p-terphenyl organic scintillator to enhance the light yield and response time. • The signals of spent-fuel bundles with low activity were observed with the new detector. [ABSTRACT FROM AUTHOR]

This paper summarizes the existing framework of safety analysis and licensing of Canada Deuterium Uranium (CANDU) reactors in Canada, established by the Canadian Nuclear Safety Commission (CNSC) and discusses challenges with implementing the lessons learned from the Fukushima accident, in the nuclear safety analysis conducted for the existing older and new CANDU reactors. This paper compares the Canadian regulatory requirements with other international nuclear regulatory guidelines, in terms of dealing with Beyond Design Basis Accidents (BDBA) and Severe Accidents, and also reviews the past and present event classification principles in safety analysis, set by the CNSC. The CNSC has published regulatory documents and technical papers outlining the challenges to be addressed with respect to implementations of the Design Extension Conditions (DEC). This paper explores the current requirements and the principles of the regulatory framework in Canada, and discusses practical approaches to determine where in every accident sequence, crediting an Emergency Mitigating Equipment (EME)would be most beneficial, for example, using a portable EME for mitigating the consequences of a Severe Accident, or preventing them. A systematic examination of CANDU Probabilistic Safety Assessment (PSA) results shows that crediting mobile EME in Level 1 PSA accident sequences have challenges and may not be the most beneficial use of EME. While adding another layer of redundancy to backup support systems (electrical power, cooling water or instrument air) further reduces the frequency of a core damage accident scenario, portable EME can only have a short-term mission time, and the main benefit of using them in Level 1 PSA, is to delay the consequences and releases. This paper proposes a thorough review of both Level 1 and Level 2 PSA models and conducting sensitivity analysis results, before making decision on where to allocate EME (i.e., in Abnormal Incident Manuals (AIM), Emergency Operating Procedures (EOP) or Severe Accident Management Guidelines (SAMG). It is crucial to balance between prevention and mitigation of severe core damage and large release frequencies to maximize the safety of the plant and minimize the impact on public safety. Instead of focusing on reducing the probability of occurrence (or frequency) of accidents by allocating every available EME to Level 1 PSA, an alternative approach would be determining where to allocate them to maintain a balance between prevention and mitigation of accidents. • Illustrative process flow-diagrams of pre and post Fukushima CANDU safety and licensing framework in Canada. • Using Probabilistic Safety Assessment (PSA) results for identification of BDBA and Design Extension Condition (DEC). • Most likely BDBA (DEC) in CANDU design are Local and Widespread Fuel Damage, with no regulatory requirement limit in Canada. • Using PSA to find balance between core-damage prevention and mitigation for allocating Emergency Mitigating Equipment (EME). • Challenges for crediting EME in CANDU's safety analysis and focusing on core-damage prevention. [ABSTRACT FROM AUTHOR]

This paper presents a new methodology using the Alternating Conditional Expectation (ACE) algorithm for reconstructing fuel pin powers in the Canada Deuterium Uranium (CANDU) reactor. Variations in assembly powers in neighboring fuel bundles are used as predictor variables to determine the response, e.g. power in each of the fuel pin. When a sufficient number of cases have been evaluated, the relation between the predictor variables and the response can be treated as a multi-variable regression problem. The ACE algorithm which is a non-parametric approach ( i.e. , no predetermined functional relationship between the predictors and the response is assumed) is then utilized for estimating optimal transformations of the predictors and the response. The geometry used is a two-dimensional, 3-by-3 fuel bundles configuration. Three hundred different configurations, which represent various power distributions, were evaluated to develop these optimal transformations. The optimal transformation is determined for each of 37 pins in a CANDU natural uranium fuel bundle. The quality of these transformations for predicting fuel pin power has been verified against an independent set of 50 two-dimensional, 3-by-3 fuel bundle configurations. The results indicate that the proposed pin power reconstruction algorithm is sufficiently accurate, giving an RMS (root-mean-squared) error of around 0.59% for all pins. The RMS error in predicting the maximum pin power is 0.60%. The development of the proposed methodology continued with the examination of its robustness. The impact of employing different training sets on the overall relative error in predicting the pin power is examined. In addition, the impact of employing different sizes of training sets is also investigated. For each training set size, a bias in predicting the maximum pin power is determined statistically based on the numerical results observed during the study. The results indicate that the methodology is likely to under-predict the maximum pin power; to circumvent this issue a multiplicative bias of 1.0118 needs to be applied, based on evaluating results from various sizes of the training sets. [ABSTRACT FROM AUTHOR]

This paper studies the effect of the drift flux model of the subchannel code ASSERT-PV on critical heat flux (CHF), void fraction and flow distribution across fuel bundles. Numerical experiments and comparison against measurements were performed to examine the trends and relative behaviour of the different components of the model under various flow conditions. The drift flux model of ASSERT-PV is composed of three components: (a) the lateral component or diversion cross-flow, caused by pressure difference between connected subchannels, (b) the turbulent diffusion component or the turbulent mixing through gaps of subchannels, caused by instantaneous turbulent fluctuations or flow oscillations, and (c) the void drift component that occurs due to the two-phase tendency toward a preferred distribution. This study shows that the drift flux model has a significant impact on CHF, void fraction and flow distribution predictions. The lateral component of the drift flux model has a stronger effect on CHF predictions than the axial component, especially for horizontal flow. Predictions of CHF, void fraction and flow distributions are most sensitive to the turbulent diffusion component of the model, followed by the void drift component. Buoyancy drift can be significant, but it does not have as much influence on CHF and flow distribution as the turbulent diffusion and void drift. [ABSTRACT FROM AUTHOR]

• Oxide wedging crack growth postulated for CANDU pressure tube manufacturing flaws. • Compressibility of oxide wedge is insignificant deviation from displacement control. • Flaw stress intensity factor effectively constant under cyclic operating stresses. • Fatigue crack growth cannot occur while oxide wedging is active mechanism. In-service inspections of Zr-2.5Nb pressure tubes in CANDU nuclear reactors have detected inner diameter surface breaking flaw indications which are classified as manufacturing flaws. It is considered prudent to evaluate potential growth by means of a postulated oxide wedging crack growth mechanism. An oxide wedging crack growth model has previously been developed and applied to predict future flaw sizes. Fatigue crack growth due to cyclic operating stresses is calculated for these flaws, and this is added to the postulated oxide wedging crack growth for fitness-for-service evaluations. This established approach is overly conservative since fatigue crack growth cannot occur while the oxide wedging crack growth mechanism is active, due to the essentially displacement-controlled loading of the flaw by the oxide wedge, which causes the stress intensity factor for the flaw to be effectively constant under cyclic operating stresses. This paper addresses this conservatism by evaluating crack growth due to postulated oxide wedging, and demonstrating that fatigue crack growth is not possible under oxide wedge loading. [ABSTRACT FROM AUTHOR]

Thermodynamic models of complex chemical systems provide an elegant and cost effective means to predict chemical interactions of materials and to provide guidance to experimental research to minimise costs. Starting in 1995 at the Royal Military College of Canada, a thermodynamic treatment of irradiated nuclear fuel has been developed that not only describes the thermochemistry of the fuel at elevated temperatures during a potential Loss-of-Coolant Accident (LOCA), but it can also be used to help predict the oxidation environment for fresh fuel measurements, irradiated fuel behaviour, the aqueous chemistry of fuel debris in coolant, nuclear waste disposal, and other systems involving nuclear fuel. Furthermore, this treatment was supported by several experimental campaigns at the Canadian Nuclear Laboratories and the Institute of Transuranium Elements for validation purposes. This paper will trace the development of this treatment and demonstrate its current practice and future potential in the general context of performance and safety of nuclear fuel. [ABSTRACT FROM AUTHOR]

The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code_Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely. [ABSTRACT FROM AUTHOR]

• This paper presents a new methodology for the qualitative screening step for CANDU reactors in Canada for performing Fire PRA. • This qualitative screening step includes a decision tree that screens out fire zones from further analysis according to pre-determined criteria. • This is very important step in any Fire PRA. • It screens out fire zones that do not affect the integrity of Fire Safe Shutdown Analysis (FSSA) cables during worst-case fire scenarios. This paper presents a new decision tree that identifies fire compartments that are screened out from further analysis according to pre-determined criteria. The paper identifies the most common fire zones and uses Computational Fluid Dynamics (CFD) model Fire Dynamics Simulator to calculate the consequences. This is a critical and very significant step in the Fire PRA. It screens out fire zones that do not affect the integrity of Fire Safe Shutdown Analysis (FSSA) cables during worst-case fire scenarios. [ABSTRACT FROM AUTHOR]

• Environmental conditions were measured during a leak rate test (LRT) of a reactor building (RB). • The environmental conditions were used to evaluate thermal strains on RB. • Internal pressure-strain relationships of four LRTs were obtained after removing thermal strains. • The RB has shown a very small decrease in the pressure-strain relationship over 18 years. With nuclear reactor power plants, Reactor Building (RB) Leak-Rate Tests (LRTs) are conducted at regular intervals to confirm the air leak tightness of a containment structure. These LRTs provide an excellent opportunity to study the containment structure's behaviour when subjected to internal pressure, thereby evaluating the structural performance of the containment structure. However, significant time-dependent fluctuations are observed in the measured strains due to the contribution of thermal strain induced by the surrounding environment. In order to accurately study the mechanical behavior of the containment structure, the thermal strain must be separated from the measured total strain. In this study, environmental data was collected during an RB LRT. Finite element analyses were carried out by utilizing the environmental conditions to predict the thermal strain of the containment structure. The predicted thermal strain was then subtracted from the measured total strain, so the underlying mechanical strain induced by the internal pressure could be approximated. The mechanical behavior of a containment structure during LRTs over an eighteen-year period is compared to understand the overall aging effects of a CANDU® 1 1 CANDU (CANada Deuterium Uranium Reactor) is a registered trademark of the Atomic Energy of Canada Limited (AECL), used under exclusive license by Candu Energy Inc. containment structure. Only a slight change in structural stiffness was observed over eighteen years though further study is suggested to confirm this. This paper outlines how field data collected during LRTs can be used to provide insight into understanding the long-term behavior of a containment structure. [ABSTRACT FROM AUTHOR]

• A improved multicell method was proposed for fuel assembly adjacent to reflector. • The proposed method has been verified for baffle-reflector and heavy-reflector PWRs. • The simulation accuracy for the power distributions can be notably improved. The conventional "two-step" scheme based on assembly homogenization is widely applied in the modeling and simulation for PWRs. In the "two-step" scheme, the reflective boundary condition is applied in the two-dimensional calculation of fuel assembly, providing the few-group constants and ADFs (Assembly Discontinuity Factors) for the three-dimensional core calculation. However, the neutron-leakage phenomenon would occur to the fuel assemblies adjacent to the reflector, in which case the approximation of the reflective boundary condition applied to the single fuel assembly would no longer be suitable. In order to consider the neutron-leakage phenomenon, the multicell method originally developed for CANDU reactor has been improved and applied to PWRs for the fuel assemblies adjacent to reflector in this paper. Through the heterogeneous simulation of the multicell case, which are composed of fuel and reflector assemblies to characterize the neutron leakage from the fuel assembly, the multicell correction factors for the few-group constants and ADFs can be determined and applied to the core calculation. Applying the improved multicell method, our home-developed Bamboo-C code has been improved. For the method verifications, the baffle-reflector PWR and heavy-reflector PWR have been modeled with the improved Bamboo-C code, and the k eff and power distributions are compared with reference values provided by the Monte-Carlo code. The numerical results show that the improved multicell method can reduce the power-distribution differences notably, which indicate the effectivity of this method for the fuel assemblies adjacent to both the baffle-reflector PWR and heavy-reflector PWR. [ABSTRACT FROM AUTHOR]

In this paper, a comprehensive vibration model is developed for analysing in-plane and out-of-plane vibration of CANDU fuel endplate rings by taking into consideration the effects of in-plane extension in the circumferential and radial directions, shear, and rotatory inertia. The model is based on Reddy’s thick plate theory and the nine-node isoparametric Lagrangian plate finite elements. Natural frequencies of various modes of vibration of circular rings obtained using the proposed method are compared with 3D finite element results, experimental data and results available in the literature. Excellent agreement was achieved. [ABSTRACT FROM AUTHOR]

This paper studied the requirements of risk monitor software and its applications as a plant specific risk monitor, which supports risk-informed configuration risk management for the two CANDU 6 units at the Third Qinshan nuclear power plant (TQNPP) in China. It also describes the regulatory prospective on risk-informed Probabilistic Safety Assessment (PSA) applications and the use of risk monitor at operating nuclear power plants, high level technical and functional requirements for the development of CANDU specific risk monitor software, and future development trends. [ABSTRACT FROM AUTHOR]

Over the past few years, Candu Energy Inc. (a wholly owned subsidiary of SNC-Lavalin Inc., which acquired the assets of Atomic Energy of Canada Limited’s Commercial Reactor Division) has been continuously developing and evaluating various options to improve the regional overpower protection (ROP) margin in aged CANDU 600 MW (CANDU 6®) reactors. This paper presents results from applying a couple of margin improvement options to a generic aged CANDU 6 reactor, namely ROP detector layout optimization and application of a revised handswitch position designation. Application of these options requires no change to the ROP analysis methodology, statistical approach or acceptance criterion. As such, any increase in ROP margin associated with these options carries little or no licensing risk and are not expected to require more than one standard outage to implement. [ABSTRACT FROM AUTHOR]

In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle. [ABSTRACT FROM AUTHOR]

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles. [ABSTRACT FROM AUTHOR]

The United Arab Emirates (UAE) is currently heavily reliant on the burning of fossil-fuels to generate most of its electric power needs. Owing to the newly set 2050 vision in combating climate change, the UAE has set an objective to considerably reduce its carbon footprint associated with power generation facilities. This vision materialized in the Barakah Nuclear Power Plant (BNPP) project, marking the first peaceful civil nuclear power plant in the Arab world. Nuclear power generates radioactive waste. In the UAE, the government's policy is to institute a safe and permanent geological disposal facility (GDF) for the final disposal of higher activity radioactive wastes (HAW). Currently, the UAE does not have a comprehensive radioactive waste inventory database on which an effective management strategy can be built on. Having a comprehensive radioactive waste data collection system will enable the UAE to not only address international reporting commitments, but also provide a transparent record of radioactive waste, which might assist in providing further public assurances. Additionally, the database will provide essential data that will support strategy development, and early waste management planning. It is not possible for the UAE to have an effective radioactive waste management strategy without implementing a system that correctly and accurately defines, records and presents radioactive waste arisings from its nuclear energy program. This paper reviews the overall radioactive waste scene in the UAE and provides a suggestion with regards to a new radioactive waste inventory management system that is based on international best practices. • Natural uranium (NU) utilization on a heavy water moderated molten salt reactor (HWMSR) is studied. • The fuel utilization efficiency of NU on HWMSR can achieve 1%, higher than that of CANDU reactor and LWR. • Around 2662 kg U-233 can be produced over 20 years' operation by online refuelingthorium and extracting Pa-233. [ABSTRACT FROM AUTHOR]

Abstract: The T-junctions are widely used in industry, such as the joint between hot leg and Automatic Depressurization System Stage Four (ADS-4)/Passive Residual Heat Removal (PRHR)/surge pipeline in AP1000 plant, header-feeder geometry of CANDU reactor, complex piping systems in the water and petroleum processing and so on. In the nuclear plant accident analysis, the T-junction entrainment is fairly closely linked with the water volume of reactor and reactor safety. Many researchers have conducted the experimental and theoretical research on the generation process, mechanism, influencing factors and modeling of entrainment. Research directions of entrainment at T-junction are multifarious, so a review work is necessary. This paper describes a comprehensive literature review of entrainment at T-junction. Based on different geometric structures of T-junction, detailed summarizations and reviews on the present entrainment phenomena, entrainment models and scaling analysis are performed. Conclusions and further research directions on entrainment at T-junction are proposed. [Copyright &y& Elsevier]

Abstract: In this paper a numerical model is developed to predict the vibration response of a CANDU 1 [1] CANDU® is a registered trade-mark of Atomic Energy of Canada Limited (AECL). fuel bundle and the associated fretting wear in the surrounding pressure tube. Two excitation mechanisms are considered in this model: turbulence-induced excitation caused by coolant flow inside the fuel channel, and seismic excitation. For the seismic excitation, simulated accelerograms of a magnitude 6 earthquake for the eastern part of Canada are considered. These accelerograms are compatible with the 2005 national building code of Canada uniform hazard spectrum (UHS). The numerical model can be easily adapted to include the effects of fuel bundle impact during the refuelling and start-up of the reactor, and the acoustic pressure pulsations caused by the primary heat transport (PHT) pumps. The simulation is performed for a typical CANDU fuel bundle with 37 fuel elements using INDAP, which is an in-house general purpose finite element program capable of simulating the nonlinear dynamics of nuclear and power equipment subjected to fluid excitation. The clearances between the buttons of the inner fuel elements, and between the bearing pads of the outer fuel elements and the pressure tube, were measured from a typical CANDU fuel bundle. Some variability among the measured clearance values was observed. Therefore, probability density functions of the measured clearance values were established and the simulation was performed for the probabilistic distribution of the clearance values. The contact between the fuel bundle and the pressure tube is modelled using pseudo-force contact method. The results, along with the proposed modelling technique, can be used in future CANDU reactors to avoid fuel and pressure tube fretting damage due to the aforementioned excitation mechanisms. [Copyright &y& Elsevier]

Abstract: 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. [Copyright &y& Elsevier]

• Probabilistic finite element analysis of CANDU fuel channels. • Computationally efficient method for probabilistic finite element analysis. • Application to 1D and 3D probabilistic contact assessment of CANDU fuel channels. • Contact assessment based on different pressure tube orientations. In CANDU reactors, irradiation induced creep and growth of the fuel channel (FC) assembly leads to Pressure Tube (PT) sag and to a reduction in the gap between the PT and the Calandria Tube (CT) (PT–CT gap). Excessive PT sag can lead to contact between the PT and the CT. Uncertainties in the parameters that influence PT–CT contact require probabilistic assessment methods to assess risk of PT-CT contact and demonstrate adherence to CSA Standard N285.8 provisions. Improved models in the prediction of PT–CT contact require computationally expensive 3D finite element analysis (3D FEA), a tool that is prohibitively difficult to use in probabilistic assessments. Current probabilistic analyses rely on 1D FEA models for the prediction of PT-CT contact estimates that are generally less accurate. Therefore, an improved and computationally more efficient approach for conducting probabilistic finite element analyses (PFEA) of CANDU FCs is highly desirable. In this paper, such a PFEA approach is proposed by coupling the multiplicative dimensional reduction method (M-DRM) and the polynomial chaos expansion (PCE) method. The analysis that follows studies the PT–CT gap predictions using deterministic 3D FEA by considering two different PT configurations present within the population of the FCs in a CANDU reactor core. The PFEA is then carried out using 1D and 3D FEA by employing the ABAQUS software suite. The accuracy of the probabilistic analysis results using the proposed approach is initially checked against Monte Carlo Simulation (MCS) results using 1D FEA. Following a good agreement, the proposed approach was applied for PFEA based on 3D FEA. The results indicate that the two different PT orientations significantly influence the probabilistic contact results. They also show that the probabilistic results based on 1D FEA, which are currently used by the nuclear industry, are not reliable. Moreover, it is also shown that a sensitivity analysis can be performed without any additional effort. The low computational cost, predictive capability and the generality of the proposed method are important attributes for adopting it in the future in probabilistic assessments of an entire reactor core. [ABSTRACT FROM AUTHOR]

• 3D model of Fuel Bundle of CANDU-6 reactor was simulated by SCALE 6.0 aiming neutronic studies of reprocessed fuel. • The Direct Use of Spent PWR Fuel in CANDU (DUPIC) was considered in the simulations. • The Oxidation and Reduction of Oxides (OREOX) method was considered to calculus of the reprocessed fuel. • The use of reprocessed fuels increases the core reactivity and so, neutron absorber materials (depleted uranium, thorium, dysprosium and hafnium) were simulated into CANDU-6 fuel bundle. Fuel economy and novel fuel cycles for nuclear industry is a wide subject that required a massive effort along the decades. The reprocessing methods has shown a great potential to reach this goal to reduce the environmental issues related to burnt fuel disposal in nuclear resources utilization frame which is the primary subject. In second to reduce the mining and milling disadvantages which has minor effect if compared with the first. Besides the advantages reprocessing wastes should be considered. Thus, advanced nuclear fuel cycles has been proposed and one of the interest for CANDU technology is the communication between CANDU and LWR fuel cycles. In this concept, for short DUPIC/OREOX method is one of the most promising due cost and lower residual products and retain non-proliferation features. Since the spent PWR fuel has a higher fissile content than natural uranium, the DUPIC fuel results in a greater amount of fissile nuclides in fuel matrix. The objective of this paper is to state a roadmap to keep neutronics of the reprocessed fuel bundle as closer as possible to natural uranium fuel bundle to achieve a reliable scenario for further CANDU 6 reactor core fueling scheme. An evaluation of neutron absorber materials (depleted uranium, thorium, dysprosium and hafnium) was presented in sense of reactivity reduction due reprocessed fuel attributes. In the systems assessment the use of hafnium had shown to be suitable relating to system criticality. The SCALE 6.0 code was used for modeling and analysis. [ABSTRACT FROM AUTHOR]

Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today''s digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed. [Copyright &y& Elsevier]

Abstract: An effective procedure based on the linear complementary problem (LCP) formulation is presented in this paper to deal with lateral contact constraints among a system of closely packed and parallel fuel rods at a cross section where spacer pads and bearing pads are introduced. The total number of potential contact constraints exceeds the number of available degrees of freedom (DOF''s). Numerical results for an example application involving deformations of seven rods at the midspan cross section indicate that the proposed procedure is accurate and robust. The scheme is currently being implemented into a computer code for computing vibration of CANDU® 2 [2] A registered trade mark of Atomic Energy of Canada Ltd. fuel bundles under normal operating conditions. [Copyright &y& Elsevier]

Abstract: For CANDU pressure tubes made from Zr–2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20–40ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term irradiation in reactor. The paper describes a prospective way for the probabilistic approach of CANDU pressure tube failure by DHC mechanisms during cool-down cycles by using probabilistic fracture mechanics principles. The limit state functions are defined for fracture instability and plastic collapse according to the Canadian Standard N285.8-05 criteria. British Procedure R6 is used to define another limit state function based on reserve factors. Discussion of lifetime probability values obtained from both procedures is made. [Copyright &y& Elsevier]

Abstract: Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors. Three different fuel compositions have been investigated: (1): 90% ThC+10% PuC, (2): 70% ThC+30% PuC and (3): 50% ThC+50% PuC. The temporal variation of the criticality k ∞ and the burn-up values of the reactor have been calculated by full power operation up to 17years. Calculated startup criticalities for these fuel modes are k ∞,0 =1.6403, 1.7228 and 1.7662, respectively. Attainable burn up values and reactor operation times without new fuel charge will be 94700, 265000 and 425000MW.D/MT and along with continuous operation periods of ∼3.5, 10 and 17years, respectively, for the corresponding modes. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically. [Copyright &y& Elsevier]

Abstract: This paper presents an experimental investigation of effects of angular misalignment on flow-induced vibration of two simulated 43-element CANDU fuel bundles in an out-reactor fresh water loop. A total of 11 different alignment configurations are examined for a range of mass flow rates from 8.56kg/s to 27.36kg/s. The dynamic response of the simulated bundles is acquired using the MEMS-based accelerometers, non-contact type displacement sensor and a data acquisition system. The experimental data show that the bundle lateral vibration is sensitive to the bundle angular misalignment; there exist few misalignment configurations which exhibit significantly large bundle lateral vibration. This is an indication that the flow in the bundle-to-bundle interface region plays an important role in flow-induced vibration of CANDU fuel bundles. [Copyright &y& Elsevier]