• Nuclear Engineering and Technology
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• 제49권1호
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• pp.6-16
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• 2017

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the $2{\times}2$ checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.642-649
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• 2016

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.875-882
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• 2014

In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU) reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.72-77
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• 2010

During the period of reinforcement work for the licensing renewal of CANDU NPP, the fuel channels, Calandria tubes and feeders of CANDU Reactor are replaced. The remote visual inspection of Calandria internal is also performed during the period of reinforcement work. This period is a unique opportunity to inspect the inside of the Calandria. The visual inspection for the Calandria vessel and its internals of Wolsong NPP Unit 1 was performed by Nuclear Engineering & Technology Institute(NETEC) of KHNP. To perform this inspection, NETEC developed equipment applied new technology such as the synchronization of 3D CAD, automatic alignment and control system. The inspection confirmed that the Calandria integrity of Wolsong NPP Unit 1 is perfect.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.175-183
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• 2004

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.267-276
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• 2017

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the $^*SIMULATE$ module of the Reactor Fueling 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. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the $^*INSTANTAN$ module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4042-4051
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• 2021

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.