• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.265-279
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• 2017

The Gyeongju underground silo type disposal facility, approved for use in December 2014, is in operation for the disposal of low and very low-level radioactive wastes, excluding intermediate-level waste. That is why the existing low-level radioactive waste level has been subdivided and the concentration limit value for intermediate-level waste has been changed in accordance with Nuclear Safety Commission Notice 2014-003. For the safe disposal of intermediate-level wastes, new optimization methodology for calculating the concentration limit of intermediate radioactive level wastes at an underground silo type disposal facility was developed. According to the developed optimization methodology, concentration limits of intermediate-level wastes were derived and the inventory of radioactive nuclides was evaluated. The operation and post closure scenarios were evaluated for the derived radioactive nuclide inventory and the results of all scenarios were confirmed to meet the regulatory limit. However, in case of $^{14}C$, it was confirmed that additional radioactivity limitation through a well scenario was needed in addition to the limit of disposal concentration. It was confirmed that the derived intermediate concentration limit of radioactive waste can be used as the intermediate-level waste concentration limit for the underground disposal facility. For the safe disposal of intermediate-level wastes, KORAD plans to acquire additional data from the radioactive waste generator and manage the cumulative radioactivity of $^{14}C$.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.562-570
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• 1995

The recent ABB/CE(Asea Brown Boveri Combustion Engineering) type pressurized oater reactor-s have the on-line monitoring system, i.e., the COLSS(core operating limit supervisory system), to prevent the specified acceptable fuel design limits from being violated during normal operation and anticipated operational occurrences. One of the main functions of COLSS is the on-line monitoring of the DNB(departure from nucleate boiling) overpower margin by calculating the MDNBR(mini-mum DNB ratio) for the measured operating condition at every second. The CETOP-D model, used in the MDNBR calculation of COLSS, is benchmarked conservatively against the TORC mod-el using an inlet flow factor of hot assembly in CETOP-D as an adjustment factor for TORC. In this study, a technique to optimize the CETOP-D inlet flow factor has been developed by elim-inating the excessive conservatism in the ABB/CE's. A correlation is introduced to account for the actual variation of the CETOP-D inlet flow factor within the core operating limits. This technique was applied to the core operating range of the YongGwang Units 3&4 Cycle 1, which results in the increase of 2％ in the DNB overpower margin at the normal operating condition, compared with that from the ABB/CE method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.143-149
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• 2012

Groundwater flow simulations were performed to obtain data of groundwater flow used in a safety assessment for a hypothetical geological disposal facility assumed to be located in the KURT (KAERI Underground Research Tunnel) site. A regional scale modeling of the groundwater flow system was carried out to make boundary conditions for a local scale modeling. And, fracture zones identified at the study site were involved in the local scale groundwater flow model. From the results of the local scale modeling, a hydraulic head distribution was indicated and it was used in a particle tracking simulation for searching pathway of groundwater from the location of the hypothetical disposal facility to the surface where the groundwater reached. The flow distance and discharge rate of the groundwater in the KURT site were calculated. It was thought that the modeling methods used in this study was available to prepare the data of groundwater flow in a safety assessment for a geological disposal facility of radioactive wastes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.83-90
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• 2020

The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the ²³⁸U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. ¹⁰B/¹¹B, ³⁶Cl/³⁵Cl, ⁴⁸Ti/⁴⁹Ti, and ²³⁵U/²³⁸U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between ⁶Li/⁷Li and plutonium production depended on the initial concentration of the impurities in graphite. Although ⁷Li can be produced through the neutron capture reaction of ⁶Li, the (n, α) reaction of ¹⁰B was the major source of ⁷Li. Therefore, the initial concentration of ¹⁰B affected the production of ⁷Li, making Li unsuitable as an indicator element for the GIRM.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.1-14
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• 2017

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with $CoCl_2$ and $K_4Fe(CN)_6$ solutions. When CHA, with average particle size of more than $10{\mu}m$, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than $10^4mL{\cdot}g^{-1}$) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.25-31
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• 2009

An analytical method of $^{99}Tc$ concentration in soil was set up and discussed considering the $^{99}Tc$ concentration in Korean soil measured with its analytical method. A selective TEVA resin was used to separate and purify the $^{99}Tc$ in the soil sample. $^{99m}Tc$ from a commercial $^{99}Mo/^{99m}Tc$ generator was used as a yield tracer for the chemical separation of $^{99}Tc$ and its problem when using $^{99m}Tc$ as a tracer was discussed. The chemical recovery yield of $^{99}Tc$ was above 70%. The optimum conditions of inductively coupled plasma mass spectrometry system(ICP-MS) were set up to determine the $^{99}Tc$ after the separation process. The minimum detectable activity(MDA) was 15 mBq/kg-dry in this analytical procedure. The $^{99}Tc$ concentration in soils of Jeju and Kori were measured in the rage of 33.73-89.16 mBq/kg-dry. Those values were less than those reported in other countries and seemed to be originated from atmospheric fallout.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analyses were performed for the container in a view of Type B. As a result of structural analysis, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at $134.8^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.103-116
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• 2006

By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.295-303
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• 2012

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advection-dispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed $10^{-3}$, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.