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Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors. [ABSTRACT FROM AUTHOR]

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]

To improve the reliability of pressurizer level measurement channels against sensor common cause failures (CCFs), a sensor fault detection scheme has been developed based on analytical redundancy in this paper. The fault detection mechanism has been transformed into statistical tests on a Kalman filter generated innovation sequence. An exponentially weighted moving average (EWMA) scheme has been developed to work with other sensor channels to ensure that the abnormality can be detected and isolated as quickly as possible. Because analytical redundancy does not share the same failure modes as existing sensor channels, the overall reliability of the measurement system has been improved. The paper provides a detailed procedure for formulating analytical redundancy, generating the corresponding innovation sequence, and detailing the design and implementation of the proposed detection scheme. Finally, the performance of the proposed scheme has been evaluated in a CANDU (Canada deuterium uranium) pressurizer to demonstrate its potential benefits. The speed of the fault detection is also evaluated against that of the average run length of the Shewhart control chart. It can be concluded that the proposed scheme is significantly more sensitive to sensor abnormalities at an earlier stage in their development. Even though level measurement channels in a pressurizer are considered, the methodology and design techniques can easily be extended to other systems/applications. [ABSTRACT FROM AUTHOR]

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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]

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]

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]

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]

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 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]

Spent nuclear fuel out of conventional light water reactors contains significant amount of even plutonium isotopes, so called reactor grade plutonium. Excellent neutron economy of Canada deuterium uranium (CANDU) reactors can further burn reactor grade plutonium, which has been used as a booster fissile fuel material in form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates incineration of nuclear waste and the prospects of exploitation of rich world thorium reserves in CANDU reactors. In the present work, the criticality calculations have been performed with 3-D geometrical modeling of a CANDU reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Four different fuel compositions have been selected for investigations: ① 95% thoria (ThO2) + 5% PuO2, ② 90% ThO2 + 10% PuO2, ③ 85% ThO2 + 15% PuO2 and ④ 80% ThO2 + 20% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the criticality k∞ and the burnup values of the reactor have been pursued by full power operation for ~10 years. Among the investigated four modes, 90% ThO2 + 10% PuO2 seems a reasonable choice. This mixed fuel would continue make possible extensive exploitation of thorium resources with respect to reactor criticality. Reactor will run with the same fuel charge for ~7 years and allow a fuel burnup ~55 GWd/t. Copyright © 2016 John Wiley & Sons, Ltd. [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]

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]

Natural circulation is widely used in nuclear reactors as a means of cooling the core, either in accident conditions or in normal operating mode. However, this three dimensional phenomenon is complex and because of its disadvantages (like low driving forces) there is need for more experiments to be done and validation of codes, in order to prove its reliability. In case of a SBO (Station Black Out) loss of forced circulation, CANDU reactors have the ability to remove heat from the fuel passively, through single-phase or two-phase thermosyphoning and intermittent buoyancy induced flow. Many aspects have been recently studied about the thermal hydraulic performance of PHWR fuel bundle, but less attention was paid to evaluation of the pressure drops at low Reynolds (Re) number. In this paper the pressure drops in a 37-element simulated CANDU fuel bundle at low Re number, for single-phase regime, were calculated using a CFD (Computational Fluid Dynamics) model. The Re number was varied between 2500 and 24000, and velocities from 0.1 m/s to 1m/s. The obtained results were compared with correlations found in literature. [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]

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]

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]

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]

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]

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.

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]

The progression of events that develop into an accident with severe fuel or core damage in the Canada deuterium uranium (CANDU) reactor is discussed. Such events involve a number of broadly common stages in which the thermal-hydraulic behavior of the reactor fuel, fuel channels, heat transport system, and a number of key process systems governs both the rate at which severely degraded cooling conditions develop and the extent of resultant damage to the reactor core. The quantification of core damage states requires the modeling of the physical phenomena that are active in these accidents, which is a focus of this paper. As discussed in this paper, unique passive features of the CANDU reactor design have a beneficial effect in that they delay the progression of severe accidents, thereby providing ample opportunity for operator actions to stabilize the plant and mitigate the consequences. It is shown that large CANDU reactors are inherently tolerant of a prolonged loss of engineered heat sinks at decay power levels. This is because two large volumes of water (the moderator and shield water) surround the reactor core and act as in situ passive heat sinks in severe accidents. This has significant impacts on severe accident management. The pressure tube reactor design precludes melting of the core at high system pressures; that is, high-pressure melt ejection is physically impossible. In the event that severe undercooling of fuel occurs at high system pressure, a pressure tube will fail well before any significant molten fuel material can accumulate. [ABSTRACT FROM AUTHOR]

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]

The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6® reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by a heavy water moderator, as with all CANDU reactors. However, ACR uses low enriched uranium fuel, compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper presents the approach used in developing two phenomena identification and ranking tables (PIRTs) for selected ACR-700 events and their results. One of the two selected events is a large loss-of-coolant accident, which is an ACR design basis event, while the other is a severe flow blockage, which is proposed to be classified as a limited core damage event (beyond design basis event). The paper outlines the design characteristics of the ACR-700 reactor that impact the PIRT process and computer code applicability, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the finalized PIRTs. The ACR-1000 reactor design is currently in detailed design at AECL, and it retains similar design features as the ACR-700 design. Although the PIRTs presented in this paper were developed for the ACR-700 design, they are generally applicable to the ACR-1000 design. [ABSTRACT FROM AUTHOR]

This paper presents a multi-cycle and multi-objective optimization method for the refueling simulation of a 713 MWe Canada deuterium uranium (CANDU-6) reactor based on a genetic algorithm, an elitism strategy and a heuristic rule. The proposed algorithm searches for the optimal refueling patterns for a single cycle that maximizes the average discharge burnup, minimizes the maximum channel power and minimizes the change in the zone controller unit water fills while satisfying the most important safety-related neutronic parameters of the reactor core. The heuristic rule generates an initial population of individuals very close to a feasible solution and it reduces the computing time of the optimization process. The multi-cycle optimization is carried out based on a single cycle refueling simulation. The proposed approach was verified by a refueling simulation of a natural uranium CANDU-6 reactor for an operation period of 6 months at an equilibrium state and compared with the experience-based automatic refueling simulation and the generalized perturbation theory. The comparison has shown that the simulation results are consistent from each other and the proposed approach is a reasonable optimization method of the refueling simulation that controls all the safety-related parameters of the reactor core during the simulation. [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]

The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. [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]

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]