• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Journal of Energy Engineering
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• 제25권4호
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• pp.124-132
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• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Journal of Korea Technology Innovation Society
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• 제16권2호
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• pp.476-505
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• 2013

Korea government has established and pursued green technology development strategy as the core of green growth, for example, withdrawal of 27 key green technologies through 'green technology research and development comprehensive plan ('09.1)' since 'low carbon green growth' was proposed as a new national development paradigm. In this study, we performed the Delphi survey of technology levels of 131 strategic product and service technologies derived from 27 key green technologies, utilizing large-scale group of green technology experts. The survey of technology level among main five nations resulted in the world's leading nation (US) versus EU (99.4%), Japan (95.3%), Korea (77.7%), China (67.1%) and Korea was ranked fourth. The technology gap between the world's leading nation (US) and Korea is 4.1 years behind EU (3.9 years) and Japan(3.1 years), but 2.1 years earlier than China. For our nation, key green technologies with high technology level are 'improved light water reactor (90.1%)', 'silicon-based solar cell (85.0%)', 'high-efficiency low-emission car (84.5%)' in order. Depending on the investment type of key green technologies, technology level is represented as short-term (85.0%), mid-term (77.3%) and long-term (71.1%) in order, indicating that lower technology level requires mid-to long-term investment and that the investment type is set appropriate.

• Journal of the Korean Society of Radiology
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• 제12권2호
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• pp.139-148
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• 2018

Assessment of NPP(Nuclear Power Plant) decommissioning cost is very important for safe decommissioning of nuclear power plants. In the United States, which has the most NPP decommissioning experience, the cost evaluation study has been conducted since the 1970s in order to decommissioning nuclear facilities. The US NRC has conducted studies on decommissioning technology, safety and cost for a variety of reactor type and nuclear installations. In the total decommissioning costs, the end of operation licenses accounted for the largest portion, followed by spent fuel management and site restoration. In case of immediate decommissioning, spent fuel management cost increased compared to delayed decommissioning, and delayed deocmmissioning increased the cost of terminating the operation license. However, in general, delayed decommissioning does not show any significant benefit as compared with immediate decommissioning. It is necessary to consider the evaluation according to the site conditions when evaluating the cost of decommissioning domestic nuclear power plants. Also, in Korea, IAEA recommendations were applied to reorganize the radioactive waste classification system. Therefore, it is necessary to develop a method to appropriately use the decommissioning data of the preceding US Nuclear Power Plant in the new classification system when estimating the amount of radioactive waste generated during decommissioning. In particular, the establishment of the evaluation methodology for the waste to be disposed of will be an important factor in securing the accuracy of the decommissioning cost. In addition, it is necessary to construct information data that can be applied to facility characteristics and work characteristics in order to evaluate the cost of demolition of domestic nuclear power plants.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.301-316
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• 1995

experimental and computational studies ore carried out to investigate the natural convection of the single phase flow in a tank with a heated horizontal plate facing downward. This is a simplified model for investigations of the influence of a core melt at the bottom of a reactor vessel on the thermal hydraulic behavior in a oater filled cavity surrounding the vessel. In this case the vessel is simulated by a hexahedron insulated box with a heated plate Horizontally mounted at the bottom of the box. The box with the heated plate is installed in a water filled hexahedron tank. Coolers are immersed in the U-type water volume between the box and the tank. Although the multicomponent flows exist more probably below the heated plate in reality, present study concentrates on the single phase flow in a first step prior to investigating the complicated multicomponent thermal hydraulic phenomena. In the present study, in order to get a better understanding for the natural convection characteristics below the heated plate, the velocity and temperature are measured by LDA(Laser Doppler Anemometry) and thermocouples, respectively. And How fields are visualized by taking pictures of the How region with suspended particles. The results show the occurrence of a very effective circulation of the fluid in the whole How area as the heater and coolers are put into operation. In the remote region below the heated plate the new is nearly stagnant, and a remarkable temperature stratification can be observed with very thin thermal boundary. Analytical predictions using the FLUTAN code show a reasonable matching of the measured velocity fields.

• Nuclear Engineering and Technology
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• 제18권4호
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• pp.273-280
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• 1986

The purpose of this study is to obtain the experimental data of the forced flow dryout heat flux in a heat generating debris bed which simulates the degraded nuclear reactor core after severe accident. An experimental investigation has been conducted of dryout heat flux in an inductively heated bed of steel particles with upward forced flow rising coolant circulation system under atmospheric pressure. The present observations were mainly focused on the effects of coolant mass flux, particle size, bed height, and coolant subcooling on the dryout heat flux The data were obtained when carbon steel particles in the size distribution 1.5, 2.5, 3.0 and 4.0 mm were placed in a 55 mm ID Pyrex glass column and inductively heated by passing radio frequency current through a multiturn work coil encircling the column. Distilled water was supplied with variation of mass flux from 0 to 3.5 kg/$\textrm{cm}^2$ s as a coolant in the tests, while the bed height was selected as 55 mm and 110 mm. Inlet temperature of coolant varied by 2$0^{\circ}C$ and 8$0^{\circ}C$. The principal results of the tests are: (1) Dryout heat flux increases with increase of upward forcing mass flux and particle size; (2) The dryout heat flux at the zero mass flux obviously depends on the Particle size as Previous studies; (3) The forced flow dryout heat flux in the shallow bed is somewhat higher than that in the deep bed,

• Applied Chemistry for Engineering
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• 제32권4호
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• pp.478-482
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• 2021

The catalytic yields of partial oxidation of methane (POM) to hydrogen over Ln(1)-Ni(5)/SBA-15 (Ln = Dy, Eu, Pr, and Tb) were investigated in a fixed bed flow reactor under atmosphere. As 1 wt% of Eu was added to Ni(5)/SBA-15 catalyst, the O1s and Si2p core electron levels of Eu(1)-Ni(5)/SBA-15 showed the chemical shift by XPS. XPS analysis also demonstrated that the atomic ratio of O1s, Ni2p_3/2, and Si2p increased to 1.284, 1.298, and 1.058, respectively, and exhibited O^-, and O^2- oxygen and metal ions such as Eu³⁺, Ni⁰, Ni²⁺, and Si⁴⁺ on the catalyst surface. The yield of hydrogen on the Eu(1)-Ni(5)/SBA-15 was 57.2%, which was better than that of Ln(1)-Ni(5)/SBA-15 (Ln = Dy, Pr, and Tb), the catalytic activity was kept steady even 25 h. As 1 wt% of Eu was added to Ni(5)/SBA-15, the oxygen vacancies caused by strong metal-support interaction (SMSI) effect due to the strong interaction between metals and carrier are made. They are resulted in increasing the dispersion of Ni⁰, and Ni²⁺ nano particles on the surface of catalyst, and are kept catalytic activity.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3874-3897
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• 2023

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High-Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next-generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds-averaged Navier-Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

• Journal of the Computational Structural Engineering Institute of Korea
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• 제37권3호
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• pp.205-216
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• 2024

Recently, seismic Probabilistic Safety Assessment (PSA) methods have been developed for process plants, such as gas plants, oil refineries, and chemical plants. The framework originated from the PSA of nuclear power plants, which aims to assess the risk of reactor core damage. The original PSA method was modified to adopt the characteristics of a process plant whose purpose is continuous operation without shutdown. Therefore, a fault tree, whose top event is shut down, was constructed and transformed into a Bayesian Network (BN), a probabilistic graph model, for efficient risk-informed decision-making. In this research, the fault tree-based BN from the previous research is further developed to consider the multi-hazard of earthquake-induced fire and explosion (EQ-induced F&E). For this purpose, an event tree describing the occurrence of fire and explosion from a release is first constructed and transformed into a BN. And then, this BN is connected to the previous BN model developed for seismic PSA. A virtual plot plan of a gas plant is introduced as a basis for the construction of the specific EQ-induced F&E BN to test the proposed BN framework. The paper demonstrates the method through two examples of risk-informed decision-making. In particular, the second example verifies how the proposed method can establish a repair and retrofit strategy when a shutdown occurs in a process plant.

• Applied Chemistry for Engineering
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• 제21권6호
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• pp.648-652
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• 2010

Catalytic activities of the partial oxidation of methane (POM) to hydrogen were investigated over Pd(5)/Ti-SPK and Pd(5)/Zr-SPK in a fixed bed flow reactor (FBFR) under atmosphere, and the catalysts were characterized by BET, XPS, XRD. The BET surface areas, pore volume and pore width of Horvath-Kawaze, micro pore area and volume of t-plot of Pd(5)/Ti-SPK and Pd(5)/Zr-SPK were $284m^2/g$, $0.233cm^3/g$, 3.9 nm, $30m^2/g$, $0.015cm^3/g$ and $396m^2/g$, $0.324cm^3/g$, 3.7nm, $119m^2/g$, $0.055cm^3/g$, repectively. The nitrogen adsorption isotherms were type IV with hysteresis. XPS showed that Si 2p and O 1s core electronlevels of Ti-SPK and Zr-SPK substituted Ti and Zr shifted to slightly lower binding energies than SPK. The oxidation states of Pd on the surface of catalysts were $Pd^0$ and $Pd^{+2}$. XRD patterns showed that crystal structures of fresh catalyst changed amorphous into crystal phase after reaction. The conversion and selectivity of POM to hydrogen over Pd(5)/Ti-SPK and Pd(5)/Zr-SPK were 77, 84% and 78, 72%, respectively, at 973 K, $CH_4/O_2$ = 2, GHSV = $8.4{\times}10^4mL/g_{cat}{\cdot}h$ and were kept constant even after 3 days in stream. These results confirm superior activity, thermal stability, and physicochemical properties of catalyst in POM to hydrogen.