• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• 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.1
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• pp.85-94
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• 2007

The Korean nuclear fuel cycle was modeled by the dynamic analysis method, which was applied to the once-through and alternative fuel cycles. First, the once-through fuel cycle was analyzed based on the Korean nuclear power plant construction plan up to 2015 and a postulated nuclear demand growth rate of zero after 2015. Second, alternative fuel cycles including the direct use of spent pressurized water reactor fuel in Canada deuterium uranium reactors (DUPIC), a sodium-cooled fast reactor and an accelerator driven system were assessed and the results were compared with those of the once-through fuel cycle. The once-through fuel cycle calculation showed that the nuclear power demand would be 25 GWe and the amount of the spent fuel will be ${\sim}65000$ tons by 2100. The alternative fuel cycle analyses showed that the spent fuel inventory could be reduced by more than 30% and 90% through the DUPIC and fast reactor fuel cycles, respectively, when compared with the once-through fuel cycle. The results of this study indicate that both spent fuel and uranium resources can be effectively managed if alternative reactor systems are timely implemented along with the existing reactors.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Nuclear Engineering and Technology
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• v.53 no.12
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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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• v.51 no.7
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• pp.1860-1865
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• 2019

The confinement degradation of the energetic particles during RMP would be a key issue in success of realizing the successful energy production using fusion plasma, because a 3.5 MeV energetic alpha particle should be able to sustain the burning plasma after the ignition. As KSTAR recent results indicate the generation of high-performance plasma(${\beta}_p{\sim}3$), the confinement of the energetic particles is also an important key aspect in neutral beam driven plasma. In general, the measured absolute value of the neutron intensity is generally used for to estimating the confinement time of energetic particles by comparing it with the theoretical value based on transport calculations. However, the availability of, but for its calculation process, many accurate diagnostic data of plasma parameters such as thermal and incident fast ion density, are essential to the calculation process. In this paper, the time evolution of the neutron signal from an He3 counter during the beam blank has permitted to facilitate the estimation of the slowing down time of energetic particles and the method is applied to investigate the fast ion effect on ELM suppressed KSTAR plasma which is heated by high energy deuterium neutral beams.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2570-2575
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• 2021

The exchange behaviors of hydrogen isotopes between protonated lithium metal compounds and deuterated water or tritiated water were investigated. The various protonated lithium metal compounds were prepared by acid treatment of lithium metal compounds with different crystal structures and metal compositions. The protonated lithium metal compounds could more effectively reduce the deuterium concentration in water compared with the corresponding pristine lithium metal compounds. The H⁺ in the protonated lithium metal compounds was speculated to be more readily exchangeable with hydrons in the aqueous solution compared with Li⁺ in the pristine lithium metal compounds, and the exchanged heavier isotopes were speculated to be more stably retained in the crystal structure compared with the light protons. When the tritiated water (157.7 kBq/kg) was reacted with the protonated lithium metal compounds, the protonated lithium manganese nickel cobalt oxide was found to adsorb and retain twice as much tritium (163.9 Bq/g) as the protonated lithium manganese oxide (69.9 Bq/g) and the protonated lithium cobalt oxide (75.1 Bq/g) in the equilibrium state.

• Biomolecules & Therapeutics
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• v.29 no.5
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• pp.527-535
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• 2021

Sclerostin (SOST), a regulator of bone formation in osteocytes, inhibits the canonical Wnt signaling by interacting with low-density lipoprotein receptor-related protein 5/6 (LRP5/6) to prevent Wnt binding. Loss-of-function mutations of the SOST gene caused massive bone outgrowth and SOST-null mouse exhibited a high bone density phenotype. Therefore, SOST has been suggested as a promising therapeutic target for osteoporosis. A few previous studies with X-ray crystallography identified the binding interfaces between LRP6 and SOST, but there are limitations in these studies as they used truncated SOST protein or SOST peptide. Here, we analyzed the conformational dynamics of SOST-LRP6 E1E2 complex using hydrogen/deuterium exchange mass spectrometry (HDX-MS). We examined the effect of the C-terminal tail of SOST on LRP6 conformation upon complex formation. HDX-MS analysis suggested a new potential binding interface for the C-terminal region of SOST that was missing from the previous crystal structure of the SOST-LRP6 E1E2 complex.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.262-268
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• 2022

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations. The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples - normal and cancerous breast tissues - including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles. Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic g-rays of the cancerous tissue over the normal tissue is approximately 2.

• Korean Membrane Journal
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• v.1 no.1
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• pp.1-7
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• 1999

Pd-ceramic composite membranes and catalytic membrane reactors(CMR) have been studied for hydrogen and its isotopes (deuterium and tritium) purification and recovery in the fusion reactor fuel cycle. Particularly a closed-loop process has been studied for recovering tritium from tritiated water by means of a CMR in which the water gas shift reaction takes place. The development of the techniques for coating micro-porous ceramic tubes with Pd and Pd/Ag thin layers is described : P composite membranes have been produced by electroless deposition (Pd/Ag film of 10-20 $\mu$m) and rolling of thin metal sheets (Pd and Pd/Ag membranes of 50-70 $\mu$m). Experimental results of the electroless membranes have shown a not complete hydrogen selectivity because of the presence of some defects(micro-holes) in the metallic thin layer. Conversely the rolled thin Pd and Pd/ag membranes have separated hydrogen from the other gases with a complete selectivity giving rise to a slightly larger (about a factor 1.7) mass transfer resistance with respect to the electroless membranes. Experimental tests have confirmed the good performances of the rolled membranes in terms of chemical stability over several weeks of operation. Therefore these rolled membranes and CMR are adequate for applications in the fusion reactor fuel cycle as well as in the industrial processes where high pure hydrogen is required (i.e. hydrocarbon reforming for fuel cell)