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Zirconium alloy (Zircaloy-4) is widely used for fuel cladding in the Canadian Deuterium Uranium (CANDU) heavy water reactors. However, degradation due to waterside corrosion can limit the in-reactor design life of the nuclear fuel. For this reason, efforts must be made to understand the mechanisms of corrosion and to mitigate its effects. The purpose of the experimental research related to this work consists of the assessment of corrosion of the CANDU fuel cladding material. The paper presents the results obtained by in-situ monitoring of the Pressurized Heavy Water Reactor (PHWR) water chemistry effect on fuel cladding corrosion, Zircaloy-4. Optical metallographic and Scanning Electron Microscopies (SEM), as well as X-ray diffraction (XRD) analysis, were used to evaluate the corrosion behavior of the fuel cladding material, Zircaloy-4 coupons, exposed in a Primary Heat Transport System (PHTS) autoclave system. The obtained results are useful for characterizing the corrosion behavior of Zircaloy-4 coupons exposed in the autoclave system at Cernavoda NPP for long periods. [ABSTRACT FROM AUTHOR]

The inspection of nuclear power plants is an essential process that occurs during plant outages. During this process, various systems are inspected, including the reactor's fuel channels to ensure that they are safe and reliable for the plant's operation. The inspection of Canada Deuterium Uranium (CANDU®) reactor pressure tubes, which are the core component of the fuel channels and house the reactor fuel bundles, is performed using Ultrasonic Testing (UT). Based on the current process that is followed by Canadian nuclear operators, the UT scans are manually examined by analysts to locate, measure, and characterize pressure tube flaws. This paper proposes solutions for the auto-detection and sizing of pressure tube flaws using two deterministic algorithms, the first uses segmented linear regression, while the second uses the average time of flight (ToF) within ±σ of µ. When compared against a manual analysis stream, the linear regression algorithm and the average ToF achieved an average depth difference of 0.0180 mm and 0.0206 mm, respectively. These results are very close to the depth difference of 0.0156 mm when comparing two manual streams. Therefore, the proposed algorithms can be adopted in production, which can lead to significant cost savings in terms of time and labor. [ABSTRACT FROM AUTHOR]

The nuclear industry is focusing some efforts on increasing the operational safety of current nuclear reactors and improving the safety of future types of reactors. In this context, the paper is focused on testing and evaluating the corrosion behavior of a thin chromium coating, deposited by Electron Beam Physical Vapor Deposition on Zy-4. After autoclaving under primary circuit conditions, the Cr-coated Zy-4 samples were characterized by gravimetric analysis, optical microscopy, SEM with EDX, and XRD. The investigation of the corrosion behavior was carried out by applying three electrochemical methods: potentiodynamic measurements, EIS, and OCP variation. A plateau appears on the weight gain evolution, and the oxidation kinetics generate a cubic oxidation law, both of which indicate a stabilization of the corrosion. By optical microscopy, it was observed a relatively uniform distribution of hydrides along the samples, in the horizontal direction. By SEM investigations it was observed that after the autoclaving period, the coatings with thickness from 2 to 3 µm are still adherent and maintain integrity. The XRD diffractograms showed a high degree of crystallinity with the intensity of chromium peaks higher than the intensity of zirconium peaks. Electrochemical results indicate better corrosion behavior after 3024 h of autoclaving. [ABSTRACT FROM AUTHOR]

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A competing-rates model is presented to account for operational changes in the metastable β-Zr phase of the Zr-2.5Nb alloy used to make CANDU reactor pressure tubes and is used to predict temperature gradients at the outlet rolled joints using the decomposition of the β-Zr phase as a proxy for temperature. High temperatures decompose the β phase by enhancing the formation of small particles of ω and α phases. Fast neutron flux causes the ω and α phases to shrink. This process is assumed to depend on the total volume of the particles, because they are comparable to, or smaller than, the size of the neutron displacement cascades. The barrier energy for thermal growth was determined to be 2.43 eV, when an Arrhenius A factor of 1013/s was assumed. The cross section for (ω+α)-phase shrinkage is 24.5 barns for Zr-2.5Nb irradiated in CANDU reactors. Assuming that the shrinkage is dominated by the migration of self-interstitial point defects, a defect production efficiency of 1.4% was found. [ABSTRACT FROM AUTHOR]

In Canada, safe operation of CANDU (CANada Deuterium Uranium; it is a registered trademark of Atomic Energy of Canada Limited) reactors is supported by a full-scope program of nuclear safety research and development (R&D) in key technical areas. Key nuclear R&D programs, facilities, and expertise are maintained in order to address the unique features of the CANDU as well as generic technology areas common to CANDU and LWR (light water reactor). This paper presents an overview of the CANDU safety R&D which includes background, drivers, current status, challenges, and future directions. This overview of the Canadian nuclear safety R&D programs includes those currently conducted by the COG (CANDU Owners Group), AECL (Atomic Energy of Canada Limited), Candu Energy Inc., and the CNSC (Canadian Nuclear Safety Commission) and by universities via UNENE (University Network of Excellence in Nuclear Engineering) sponsorship. In particular, the nuclear safety R&D program related to the emerging CANDU ageing issues is discussed. The paper concludes by identifying directions for the future nuclear safety R&D. [ABSTRACT FROM AUTHOR]

European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used) is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA). In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the accident progression consequences up to a certain point. The code assesses the fuel bundle damage progression, but cannot assess further core damage for a CANDU type core, and starting from these data the amount of damaged fuel can be calculated. The radio nuclei present in the damaged fuel are supposed to be released into the main heat transport system and after that into the containment building in the worst case scenario. Assessing the radioactive nuclei maximum release is the purpose of the present paper. The radioactive nuclei release is needed for the accident management plan, limiting the environmental and population impact of the supposed accident, and furthermore for a later site remediation plan that can be put in action after the complete mitigation of the accident consequences. The maximum quantity of radio nuclei released during the accident calculated in this paper is a worst case scenario evaluation that can lead to better preparedness in an accident scenario. [ABSTRACT FROM AUTHOR]

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable. [ABSTRACT FROM AUTHOR]

The authors contribution to this paper is to present a concept solution of a remote control robot (RCR) used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF), unblock and extract the channel shield plug (from Lattice Tube - LT), cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR) represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection. [ABSTRACT FROM AUTHOR]

The authors contribution to this paper is to present a concept solution of a remote control robot (RCR) used for the horizontal fuel channels pressure tube decommissioning in the CANDU nuclear reactor. The authors highlight in this paper, few details of geometry, operations, constraints by kinematics and dynamics of the robot movement outside of the reactor fuel channel. Outside operations performed has as the main steps of dismantling process the followings: positioning front of Calandria structure at the fuel channel to be decommissioned, coupling and locking to the End Fitting (EF), sorting and storage extracted items in the safe container. All steps are performed in automatic mode. The remote control robot (RCR) represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the outside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection. [ABSTRACT FROM AUTHOR]

The super multi-functional calculation program for nuclear design and safety evaluation is a general, intelligent, accurate and precise simulation software system for the nuclear design and safety evaluations. The heavy water reactor has a much stronger moderation power and much longer diffusion length of the thermalized neutrons. The paper intends to show the verification and validation of SuperMC3.2 with a heavy-water-moderated lattice named the deuterium critical assembly which is very similar to the Canada Deuterium Uranium type reactor and selected from the international reactor physics experiment evaluation project. The calculation results were compared with the reference calculated results and the experimental data from International Reactor Physics Experiment Evaluation Project. The final obtained results proved the accuracy, convenience and universality of SuperMC, and primarily verified the applicability of SuperMC in nuclear analysis of heavy water reactor. [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]

This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP) by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data froma real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students. [ABSTRACT FROM AUTHOR]

An introduction is presented in which the editor discusses various reports published within the issue on topics including radiation safety, CANDU nuclear power plant, and nuclear reactor designs.

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects. [ABSTRACT FROM AUTHOR]

In CANDU reactors, shim operation is used when the online refuelling capability becomes temporarily unavailable. Adjuster rods, normally in-core to provide flux flattening, are withdrawn in sequence to provide additional reactivity as the fuel is depleted. In a CANDU 900 reactor, up to three of the eight adjuster banks may be withdrawn, with the power derated accordingly. In this study, the shim operation was modelled using a combination of TRACE_Mac1.1, PARCS_Mac1.1, and scripts modelling the reactor regulating system, all running as a single coupled simulation. A driver script simulated the operation as a sequence of steady-state, depletion, and transient models. The results were compared to operational data from a nuclear power plant, evaluating the key figures of merit. The simulation was extended beyond the operational data by reducing the power to 59% FP and withdrawing the remaining adjusters, to observe the behaviour of the simulated reactor for a deeper reactivity-driven transient. Sensitivity cases, including adjuster rod depletion and nuclear data uncertainty, were also evaluated. This study was able to successfully reproduce the general results of the shim operation. Some discrepancies were observed between the simulation and dataset for the duration of the shim, particularly for the one bank out phase of the shim. Several potential causes for the early phase behaviour were identified. When the simulation was extended, the model predicted that a power reduction below 60% FP would lead to xenon poison out when the adjusters were depleted, with the timing sensitive to the adjuster depletion. Nodalisation of the PARCS model also had a significant impact, due to the effect on adjuster nodalisation and its area-of-effect with respect to the actual adjuster locations. Nuclear data uncertainty had a lesser but still noticeable effect. Other parameters, such as the distribution of fuel burnups in the core, only had a small effect on the shim operation. [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]

• 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]

Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical. [ABSTRACT FROM AUTHOR]

This paper presents some details of operation process for a Cutting and Extraction Device (CED) in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED) is his capability of totally operator's protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator), a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs). The Cutting and Extraction Device (CED) is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC) and controlled by an operator panel type Human Machine Interface (HMI). All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons. [ABSTRACT FROM AUTHOR]

This paper presents some details of operation process for a Cutting and Extraction Device (CED) in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED) is his capability of totally operator's protection against the nuclear radiation during pressure tube decommissioning. The movement and fixing processes present few particularities due to special adopted technical solutions: train guiding-fixing modules equipped with elastic guiding rollers and fixing claws, traction modules with elastic rollers and variable pitch, also with propriety to adapt the system according to various dimensions of the tube. The Cutting and Extraction Device (CED) is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC) and controlled by an operator panel type Human Machine Interface (HMI). All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons. [ABSTRACT FROM AUTHOR]

The scope of this paper is to achieve the device functioning steps for the decommissioning of the horizontal fuel channels of calandria vessel. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. For the decommissioning operation design shall be taken to ensure all aspects of security, environmental protection during decommissioning operation steps and creating and implementing work procedures resulting from developed decommissioning plan. The fuel channel decommissioning device is designed for dismantling and extraction of the fuel channel and its components. The decommissioning operation consists of following major steps: platform with device positioning to the fuel channel to be dismantled; coupling and locking the device at the fuel channel; unblock, extract and store the channel closure plug; unblock, extract and store the channel shield plug; block and cut the middle and the end of the pressure tube; block, extract and store the end fitting; block, extract and store the half of pressure tube; mounting of the extended closing plug. The operations steps are performed by the Cutting and Extraction Device and by the extraction actuator from the device handling elements assembly. After each step of dismantling is necessary the confirmation its finalization in order to perform the next operation step. The dismantling operation steps of the fuel channel components are repeated for all the 380 channels of the reactor, from the front of calandria side (plane R) as well as the rear side (plane R'). [ABSTRACT FROM AUTHOR]

This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED) is fully automated, connected by wires to a Programmable Logic Controller (PLC) and controlled from a Human Machine Interface (HMI). The Cutting and Extraction Device (CED) performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF), unblock and extract the channel shield plug (from Lattice Tube - LT), block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next action after confirming the action in progress. The design of the Cutting and Extraction Device (CED) shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers. [ABSTRACT FROM AUTHOR]

This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI) phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6. [ABSTRACT FROM AUTHOR]

The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analysis, this paper proposes utilizing a B1 theory augmented Monte Carlo (MC) few-group constant generation method (B1 MC method) which has been devised for the PWR fuel assembly analysis method. To examine the applicability of the B1 MC method for the CANDU 6 core analysis, the fuel bundle cell and supercell calculations are performed using it to obtain the two-group constants. By showing that the two-group constants from the B1 MC method agree well with those from WIMS-AECL and that core neutronics calculations for hypothetical CANDU 6 cores by a deterministic diffusion theory code SCAN with B1 MC method generated two-group constants also agree well with whole core MC analyses, it is concluded that the B1 MC method is well qualified for both fuel bundle cell and supercell analyses. [ABSTRACT FROM AUTHOR]

The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling. [ABSTRACT FROM AUTHOR]

This paper looks at the existing challenges with steady-state Liquid Zone control at some CANDU (CANada Deuterium Uranium) stations, where--contrary to expectations for equilibriumflow--Liquid ZoneControlValve oscillations have proven to be a chronic, unanticipated challenge. Currently, the exact causes of this behaviour are not fully understood, although it is confirmed that the Control Valve oscillations are not due to automatic power adjustment requests or zone level changes due to process leaks. This phenomenon was analysed based on a case study of one domestic nuclear power station to determine whether it could be attributed to inherent controller properties. Next, a proposal is made in an attempt to improve current performance with minimal changes to the existing systemhardware and logic using conventional technologies. Finally, a proposal wasmade to considerModel Predictive Control-based technology to minimize the undesirable Control Valve oscillations at steady state based on the obtained simulation results and discussion of other available alternatives. [ABSTRACT FROM AUTHOR]

Abstract: A theoretical analysis has been developed to predict the critical height of the onset of gas entrainment (OGE) during dual and triple discharge from a stratified two-phase region. The two and three discharge branches are mounted on a circular wall, resembling a circular reservoir of a CANDU header–feeder configuration. A point sink model has been developed to predict the critical height and to map the velocities and acceleration flow fields during OGE. The model was verified by comparing the theoretically predicted critical height with the available experimental results. The theoretically predicted critical height is found to be a function of the branch Froude number, the location of the secondary branch with respect to the primary branch, and the angle between the branches. The effect of these variables on the predicted OGE height was investigated and is presented in this paper. Predictions of the critical height were shown to be within 25% of experimental values in both dual and triple discharge. [Copyright &y& Elsevier]

Delayed hydride cracking (DHC) in Zr–2.5Nb alloy material is of interest to the CANadian Deuterium Uranium (CANDU)(1) industry in the context of the potential to initiate DHC at a blunt flaw in a CANDU reactor pressure tube. The existing CANDU blunt flaw DHC evaluation procedure is based on a threshold peak flaw-tip stress for DHC initiation that is independent of flaw geometry. Work is underway to improve the existing blunt flaw DHC evaluation procedure by developing a methodology that takes into account the effect of flaw geometry parameters. The methodology is based on representing the stress relaxation due to hydride formation and crack initiation by a process zone. A process-zone model was used to develop failure assessment diagrams in the conventional format of R6 as described in this paper, although the coordinates of the diagram have a different physical meaning. The failure assessment diagrams were shown to be independent of the scale of the configuration as well as specific values of material properties. The failure assessment curves as derived in the conventional R6 format exhibit a strong dependence on the geometry of the flaw due to the stress concentration and associated stress gradient. By redefining the ordinate of the failure assessment diagram in terms of peak flaw-tip stress the geometry dependence of the failure assessment curves was reduced significantly. This minimal geometry dependence of the failure assessment curves is valuable with regards to practical engineering flaw evaluations. Agreement between the failure assessment diagram predictions and experimental results is reasonable. [Copyright &y& Elsevier]

• ISF with four distillation columns cascade to separate the deuterium-tritium mixture. • Control systems: Plant Control System (SCADA) and Safety Instrumented System (SIS). • SCADA – keep the process within the accepted limits - 1st layer of defence in depth. • SIS – performs the assigned security functions - 2nd defence layer in depth. Isotopes separation technologies are used in both Tritium Removal Facilities associated with CANDU reactors (to remove tritium from heavy water) and fusion applications (eq. ITER Isotope Separation System) to recover and enrich tritium for further operation. The expertise gained during design, manufacturing, building, commissioning and operation of the Cryogenic Distillation System (CDS) with proprietary four column design built at ICSI Rm. Valcea, could be applied for various tritium recovery applications, regardless of the type of front-end process used, Liquid Phase Catalytic Exchange or Combined Electrolysis and Catalytic Exchange. Cryogenic Distillation feed is a mixture of hydrogen isotopes which presents hazards in operation (explosion and tritium contamination), therefore the CDS of PESTD was designed to meet industrial and nuclear applicable standards. In this paper we present an overview of the CDS focusing on the process control and safety approach which meet specific conditions of Romanian codes and standards required for operation. However, specific information may be used later as reference for design consideration for other CDS. The cryogenic distillation process from ICSI Rm. Valcea is designed as four columns cascade inside a cold-box which acts as second barrier against hydrogen release and thermal insulation. The CDS process control is based on a Programmable Automation Controller (PACs), a Supervisory Control and Data Acquisition (SCADA) and specific low temperature field equipment and devices. The control room operator supervises the process using a specific designed Human Machine Interface (HMI) which permits to monitor process parameters, alarms acknowledgment and acts to adjust the process or override normal PACs control. [ABSTRACT FROM AUTHOR]

• Fluctuation of strain during LRT of a containment structure is due to temperature. • A method has developed to run thermo-mechanical analysis of a containment structure. • Strain history from the numerical analysis agrees well with field measurements. • The proposed method can be potentially used to remove noise in measured strain. A strategy for isolating the thermal strain due to solar radiation from the total measured strain during the Integrated Leak Rate Test of a nuclear containment structure is presented in this paper. In the proposed method, a series of numerical models are developed to estimate the temperature change on the prestressed concrete containment structure and to predict the strain due to temperature and applied pressure. The effects of the solar radiation, heat exchange with internal and external air and heat conduction are considered in the proposed method. The developed method is validated by reproducing the total strain profiles measured from the leak rate tests conducted at the Point Lepreau 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. 6 containment structure. [ABSTRACT FROM AUTHOR]

A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF) of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle. [ABSTRACT FROM AUTHOR]

CFD simulations of cross-flows along in-line and staggered tube bundles which emulate those encountered in the calandria of CANDU-6 reactors are presented. The knowledge of external wall temperature distributions around calandria tubes is a major concern during normal and off-normal operating conditions of CANDU reactors. Calculations are performed using the FLUENT software with several turbulence models using segregated and Coupled algorithms. It is observed that κ-based models are able to reproduce mean velocities in staggered bundles. In most cases, the Coupled algorithm yields convergence even if it requires a longer computational time. Based on this work, the standard κ-ε model is recommended to perform this kind of simulations. Improved κ-ε models do not lead to better results while the κ-ω model predicts very well the physics only around the first row but it is unable to predict the flow around tubes located far downstream in the bundle. [ABSTRACT FROM AUTHOR]

In delayed hydride cracking (DHC) of Zr-2.5wt% Nb alloy, the hydride cluster at the crack tip has a critical length, which is a function of the stress intensity factor KI and other parameters. When KI > KIH, the threshold stress intensity factor, the hydride cluster must grow to this critical length before it will fracture. On the other hand, when KI < KIH, there is a maximum length to which the hydride cluster can grow, and this length is insufficient for fracture i.e. less than the critical length. In this work, the lengths of the hydride cluster were experimentally studied for KI < KIH and KI > KIH near KIH. A modified experimental method was used, that permitted the hydride clusters to be formed and fractured individually. The hydride clusters were observed to be wedge-shaped, in agreement with the predictions by Metzger and Sauvé (PVP, vol. 326, ASME, 1996). The lengths of hydride cluster measured in this work are compared with existing theoretical predictions. A good general agreement was obtained, but some differences are discussed. [ABSTRACT FROM AUTHOR]