• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.357-365
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• 2016

We observed the ${\gamma}-ray$ shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of $0.371cm^{-1}$ from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a ${\gamma}-ray$ of $^{137}Cs$, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of $3,175kg{\cdot}m^{-3}$. Although the unit weight of the concrete with OSS and OSG was $3,052kg{\cdot}m^{-3}$, which was lower than the maximum unit weight condition by $123kg{\cdot}m^{-3}$, its attenuation coefficient was improved by $0.012cm^{-1}$. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced ${\gamma}-ray$ shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing.

• 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.18 no.1
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• pp.51-62
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• 2020

This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to ¹³⁷Cs, ⁶⁰Co, ¹⁵²Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.63-72
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• 2020

Dissolved hexavalent uranium can exist in the form of several different chemical species. Furthermore, species distributions depend on the pH value of the aqueous solution. Representatively, UO₂²⁺, UO₂OH⁺, (UO₂)₂(OH)₂²⁺, and (UO₂)₃(OH)₅⁺ species coexist in solutions at acidic and circumneutral pH values. When amorphous silica particles are suspended in an aqueous solution, the dissolved chemical species are easily adsorbed onto silica surfaces. In this study, it was examined whether the species distribution of the adsorbed U(VI) on a silica surface followed that of the dissolved U(VI) in an aqueous solution. Time-resolved luminescence spectra of three different dissolved species (UO₂²⁺, UO₂OH⁺, and (UO₂)₃(OH)₅⁺) and two different adsorbed species (≡SiO₂UO₂, ≡SiO₂(UO₂)OH^-, or ≡SiO₂(UO₂)₃(OH)^5-) were measured in the pH range 3.5-7.5. The spectral shapes of these chemical species were compared by changing the pH value; consequently, it was confirmed that the species distribution of the adsorbed U(VI) species was different from that of the dissolved U(VI) species.

• 한국지구물리탐사학회:학술대회논문집
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• 2000.09a
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• pp.54-69
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• 2000

There are groundwater trouble by high-salinity yield inducing sea-water intrusion in Cheju Island. It is used groundwater-GIS(Well-lnfo) in the maintenance and management of groundwater in Cheju Island to grasp groundwater trouble area and cause of high-salinity yield. For 16 wells certain to yield high-salinity, we logged specific electrical conductivity(EC) and tried to get hold of freshwater and saltwater relationship. As result of distribution of $Cl^-$ by depth, it is showed up groundwater trouble by high-salinity yield in the east coastal area and the partly north coastal area. The reason of high-salinity groundwater yield are low-groundwater level by the structure of geology and low-hydraulic gradient etc. There is necessity for management to development and use of groundwater in the high-salinity area, special management area.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.271-278
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• 2004

This paper shows our experiment is performed to understand the exposure tendency of $^{137}Cs$ according to the height of area and also, to supplement it by considering chemical characters of $^{137}Cs$ exposed to the soil. The samples we use for this experiment are from the general flat area of Yeonggwang county where it has NPPs, the high places of Keumjung & Bulgap mountains, and Naejan mountain where it is quite far from the NPPs. The data from this experiment show that the exposure of $^{137}Cs$ is not harmful since its range is around 252 Bq/kg-dry in most of sampled soils such as from the general flat area, the high place of Keumjung mountain where is 2 km away from the NPPs, the other high place of Bulgap mountain where is about 20 km away from the NPPs, and Naejan mountain where it is far from the NPPs. Not like the general flat area, however, the data show that the higher the area is the more $^{137}Cs$ is exposed. That is, at the top of mountains, the more $^{137}Cs$ is exposed compared to at the bottom area. It is almost $2{\~}6$ times more than the general flat area of Yeonggwang county where it has NPPs. The data also show that the spread of $^{137}Cs$ is deeply related to the geographical(the height of area, rainfall, etc..) factors and chemical factors of soils. As the geographical factors, there are far more chances to be exposed of $^{137}Cs$ at the high area of mountains through the air compared to at lower area and therefore, we can get more high-leveled readings of $^{137}Cs$ at the high area while it is low-leveled ones at the general flat area even if both of them have the same soil conditions. Regarding the chemical factors of soil, it is clarified that the CEC is the key factor. The CEC means the capability of sticking $^{137}Cs$ accumulated into the soil. Hence, the more CEC it has the more high-leveled readings of $^{137}Cs$ we get under the same geographical condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.9-17
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• 2010

The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.

• 한국지구물리탐사학회:학술대회논문집
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• 1999.08a
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• pp.137-155
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• 1999

Recently, Televiewer(Borehole Acoustic Scanner(Televiewer)) has come to be widely used specially for the general engineering construction design. The Televiewer tool using a focussed acoustic beam is to detect the amplitude and traveltime of each reflected acoustic signal at the wall, resulting in the amplitude- and traveltime image respectively. Fractures can be well detected, because they easily scatter the acoustic energy due to the highly narrow beam. In addition, the drilling work will rough the borehole wall so that the acoustic energy can be scattered simply due to the roughness of the wall. Thus, the amplitude level can be directed associated with the elastic properties(impedance) and the hardness of the rock as well. Meanwhile, the traveltime image provides an information about the borehole shape and can be converted to a high precision 3D caliper log(max. 288 arms). In this paper, based on the high resolution of Televiewer images, general evaluation methods are illustrated to derive very reliable rock parameters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.93-100
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• 2009

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.