• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.877-887
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• 2022

A new reactor concept is described that directly couples a supercritical CO₂ (sCO₂) power cycle with a CO₂-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO₂ concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (~20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2013.05a
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• pp.63-64
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• 2013

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.101-112
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• 1995

In order to improve the recovery efficiency of heavy water vapour from the atmosphere inside a reactor building, and to recover and upgrade the heavy water which escape, special treatments, such as reducing the ingress of light water vapour, are studied in the design of the CANDU reactor systems. This is considered in controlled method of the humidity over drawing fresh air through a desiccant dehumidifier which dries the air by absorption. Comparing with the moisture loads between summer and winter operations, the moisture removal rates are calculated. Those are proportional to the difference between the controlled space and the surrounding environment Installation of a new dehumidifier will be able to reduce the moisture loads from the cooling systems, improving overall system efficiency and saving operating costs.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.1-11
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• 1990

Heavy water is used as a moderator and a coolant in the pressurized heavy water reactor Because of the high cost of heavy water, downgraded heavy water generated in the reactor system is recycled to the reactor after being concentrated up to 99.8% or more in heavy water upgraders. This study investigates the process of upgraders and then suggests a theoretical model. The relations between process variables are derived from tower packing characteristics, vapour-liquid equilibria, and mass-heat balance equations at a steady state operation of the upgrader h computer program UPGR is developed, using the algorithm that solves the nonlinear equations step by step. It shows that the results of computer simulation are in good agreement with the operating data of the Wolsung upgrader. Thus, this computer code offers the optimum operating guide and is now applied to manage the performance of upgraders for the effective operation of the heavy water upgraders.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.443-448
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• 2006

The OPAL (Open Pool Australian Light-water) reactor is currently being constructed to replace HIFAR (HI-Flux Australian Reactor, commissioned in 1958) in mid-2006. HIFAR will be shutdown for decommissioning after several months of simultaneous operation with OPAL for smooth transition of operating systems and business. OPAL is a 20 MW multipurpose research reactor for radioisotope production, irradiation services and neutron beam research. The OPAL reactor uses low enriched uranium fuel in a compact core, cooled by light water and moderated by heavy water, yielding maximum thermal flux not less than $4{\times}10^{14}ncm^{-2}s^{-1}$. The reactor containment building is constructed of reinforced concrete and has been designed to protect the reactor from all external events such as seismic occurrences and impact from a hypothetical light aircraft crash. This paper describes the main elements of the reactor design and its applications.

• Tribology and Lubricants
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• v.27 no.5
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• pp.240-248
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• 2011

Mechanical face seal installed in primary heat transport pump used for heavy water reactor prevents leakage of working fluid using thin working fluid film between primary seal ring and mating ring. If the leakage of working fluid exceeds the allowable volume, serious accident can be happened by the trouble of primary heat transport pump. The thinner fluid film exists between primary seal ring and mating ring, the less working fluid leaks out. On the other hand, if the thickness of fluid film is not enough, the life of mechanical face seal will be reduced by friction and wear. Therefore appropriate design is necessary to maximize the performance and life of mechanical face seal. In this study, numerical analysis using finite volume method was conducted to investigate the performance of mechanical face seals which have same deep straight groove and 11 different net coning values. As results, equilibrium clearance between primary seal ring and mating ring, leakage volume of working fluid, friction torque on sealing surface and stiffness of working fluid film were obtained. With increasing net coning value, equilibrium clearance and leakage volume increase, and friction torque and stiffness of fluid film decrease.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.315-322
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• 2015

The research carried out on thorium-based fuels indicates that these fuels can be considered as economic alternatives with improved physical properties and proliferation resistance issues. In the current study, neutronic assessment of $UO_2$ in comparison with two $(Th-^{233}U)O_2$, and $(Th-^{235}U)O_2$ thorium-based fuel loads in a heavy water research reactor has been proposed. The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with $UO_2$ fuel loading. By contrast, $^{235}U$-containing thorium-based fuel and $^{235}U$-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to $UO_2$ fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.