• Nuclear Engineering and Technology
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• v.13 no.3
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• pp.121-129
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• 1981

Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from the mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than tile mass leakage rate formula of CONTEMPT-LT.

• Journal of the Korean Ceramic Society
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• v.45 no.4
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• pp.245-249
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• 2008

The SiC and ZrC are critical and essential materials in TRISO coated fuel particles since they act as protective layers against diffusion of metallic and gaseous fission products and provides mechanical strength for the fuel particle. However, SiC and ZrC have critical disadvantage that SiC loses chemical integrity by thermal dissociation at high temperature and mechanical properties of ZrC are weaker than SiC. In order to complement these problems, we made new combinations of the coating layers that the ZrC layers composed of SiC. In this study, after Silicon carbide(SiC) were chemically vapor deposited on graphite substrate, Zirconium carbide(ZrC) were deposited on SiC/graphite substrate by using Zr reaction technology with Zr sponge materials. The different morphologies of sub-deposited SiC layers were correlated with microstructure, chemical composition and mechanical properties of deposited ZrC films. Relationships between deposition pressure and microstructure of deposited ZrC films were discussed. The deposited ZrC films on SiC of faceted structure with smaller grain size has better mechanical properties than deposited ZrC on another structure due to surface growth trend and microstructure of sub-deposited layer.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.104-104
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• 2018

Kori Unit 1 is the first permanent shutdown nuclear power plant in Korea and it is on June 18th, 2017. Spent fuel assemblies began to be discharged from the reactor core to the spent fuel pool(SFP) within one week after shutdown of Kori unit 1 and the campaign was completed on June 27th, 2017. The total number of spent nuclear fuel assemblies in SFP of Kori Unit-1 is 485 and their discharging date is different respectively. So, decay heat was evaluated considering the actual enrichment, operation history and cooling time of the spent fuel assemblies stored in SFP of the Kori Unit-1. The code used in the evaluation is the ORIGEN-based CAREPOOL system developed by KHNP. Decay heat calculation of PWR fuel is based on ANSI/ANS 5.1-2005, "Decay heat power in light water reactors" and ISO-10645, "Nuclear energy - Light water reactors - Calculation of the decay heat power in nuclear fuels. Also, we considered the contribution of fission products, actinide nuclides, neutron capture and radioactive material in decay heat calculation. CAREPOOL system calculates the individual and total decay heat of all of the spent fuel assemblies in SFP of Kori Unit-1. As a result, the total decay heat generated in SFP on June 28th, 2017 when the spent fuel assemblies were discharged from the reactor core, is estimated to be about 4,185.8 kw and to be about 609.5 kw on September 1st, 2018. It was also estimated that 119.6 kw is generated in 2050 when it is 32 years after the permanent shutdown. Figure 1 shows the trend of total decay heat in SFP of Kori Unit-1.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2104-2125
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• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
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• v.10 no.3
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• pp.153-163
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• 1978

The problems of solid waste disposal into the ground in connection with environmental aspects in the vicinity of a site would be very significant, though ground disposal for solid waste is safe and economical method. Studies of the waste-movement and migration of radionuclides (Sr-90 and Cs-137) for the disposal into the ground were performed under laboratory and field conditions. Affinity of the soils for radionuclide solution was higher than that in the acid solution. The sorption of radionuclides by the soils showed a time-dependent reation. The migration rates of radiostrontium and radiocesium were a range of 3.73$\times$10$^{-3}$ to 10.9$\times$10$^{-3}$ cm/day. The nuclides in the soil migrate much more slowly than the water, probably due to its high exchange capacity. The observed distribution of tritium was compared with that calculated by a mathematical model based on diffusivity. This study suggests that the tritiated water can be used to trace the movement of ground water.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.237-247
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• 1993

Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio $^{l44}$Ce $^{l37}$Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products $^{l44}$Ce and $^{l37}$Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio $^{l44}$Ce $^{l37}$Cs agreed well with the operator declared cooling time within relative difference of $\pm$5 % despite the low counting rate of the gamma-ray of $^{l44}$Ce (about 10$^{-3}$ count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of $\pm$0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio $^{l44}$Ce $^{l37}$Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.

• Korean Chemical Engineering Research
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• v.52 no.3
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• pp.279-288
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• 2014

Nuclear energy is expected to meet the growing energy demand while avoiding CO2 emission. However, the problem of accumulating spent fuel from current nuclear power plants which is mainly composed of uranium oxides should be addressed. One of the most practical solutions is to reduce the spent oxide fuel and recycle it. Next-generation fuel cycles demand innovative features such as a reduction of the environmental load, improved safety, efficient recycling of resources, and feasible economics. Pyroprocessing based on molten salt electrolysis is one of the key technologies for reducing the amount of spent nuclear fuel and destroying toxic waste products, such as the long-life fission products. The oxide reduction process based on the electrochemical reduction in a LiCl-$Li_2O$ electrolyte has been developed for the volume reduction of PWR (Pressurized Water Reactor) spent fuels and for providing metal feeds for the electrorefining process. To speed up the electrochemical reduction process, the influences of the feed form for the cathode and the type of anode shroud on the reduction rate were investigated.