• Nuclear Engineering and Technology
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• v.49 no.1
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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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• v.39 no.4
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• pp.273-286
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• 2007

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.149-149
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• 1999

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Proceedings of the Korea Association of Crystal Growth Conference
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• 1997.06a
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• pp.53-60
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• 1997

The roasting of pyrite with a cyclone reactor have been studied in terms of investigating the reaction behavior of pyrite. The development of a fundamental model for pyrite oxidation and lime sulfation in a vertical cyclone reactor. The model assumes a chemical control shrinking core behavior for the pyrite and a fluid film control shrinking core behavior for the lime. The oxygen and sulphur dioxide concentrations and the energy balance for the gas, pyrite and lime particles are solved. The model was solved and characterized numerically. Experiments have been performed to study the influence of reaction parameters such as reactor temperatures, pyrite particle sizes, air flow rates, feeding rates, and mixing ratio of pyrite and lime. The oxidation and sulfation products were characterized chemically and physically.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2002.02a
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• pp.322-325
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• 2002

In this paper, the power-linking device connecting the high-Tc super-conducting(HTS) coil to the power system in the DC reactor type three-phase high-Tc superconducting fault current limiter (SFCL) has been designed. This design was triggered from the concept that the magnetic energy could be exchanged into the electrical energy each other. Ferromagnetic material is used as the path of magnetic flux. The device mentioned above was named Magnetic Core Reactor(MCR). MCR was designed to minimize the voltage drop caused by copper loss. The current density of the conductor was 1.3 A/mm$^2$ and % voltage drop was 2%.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.647-652
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• 1998

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identified the possible even scenarios following the loss of shutdown cooling. The Thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior, From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determined the containment closure time to prevent the uncontrolled released of fission products to atmosphere, These data provide useful information to the abnormal procedure to cope with event.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.299-304
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• 1997

A neuro-fuzzy control algorithm is applied for xenon spatial oscillations in a pressurized water reactor. The consequent and antecedent parameters of the fuzzy rules are tuned by the gradient descent mettled. The reactor model used for computer simulations is a two-point xenon oscillation model. The reactor core is axially divided into two regions and each region has one input and one output and is coupled with the other region. The interaction between the regions of the reactor core is treated by a decoupling scheme. This proposed control of mettled exhibits very fast responses to a step or a ramp change of target axial offset without any residual flux oscillations.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1626-1627
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• 2007

This paper deals with the design and manufacturing of power cabinet for reactor power control system(PCS). The PCS provides the control signals and motive power to operate the CEDMs(Control Element Drive Mechanism). The CEDM is raise and lower the CEAs(Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.565-578
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• 2022

Among Generation IV reactors, the sodium-cooled fast reactor (SFR) is attracting attention as a system having great potential for commercial use. Gas entrainment is a thermal-hydraulic issue related to the safety problem of the reactor core in the SFR. Typically, a dipped plate or baffles are installed under the free surface to suppress gas entrainment. However, these approaches can cause gas entrainment in other locations and require many trial-and-error and verifications. In this study, a new strategy using magnetohydrodynamics to suppress gas entrainment in the SFR is proposed. In a counter-flow model, a judgment criterion of gas entrainment occurrence was developed for both water and liquid metal. Moreover, the gas entrainment can be completely suppressed by applying a magnetic field.