• Journal of the Korean Ceramic Society
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• v.38 no.9
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• pp.839-845
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• 2001

In order to vitrify the Ion-Exchange Resin(IER), Dry Active Waste(DAW), and borate concentrate generated from the commercial nuclear facilities, the glass formulation study based on the their compositions was performed. Two glasses named as RG-1 and DG-1 were formulated as the candidate glasses for the vitrification of hte IER and DAW, respectively. A glass named as MG-1 was also formulated as a candidate glass for the vitrification of the mixed wastes containing the IER, DAW, and borate concentrate. The process parameters, product qualities, and economics were evaluated for the candidate glasses and confirmed experimentally for the some properties. The glass viscosity and electrical conductivity as the process parameters were in the desired ranges. the product qualities such as glass density, chemical durability, phase stability, etc. were satisfactory. In case of vitrifying the wastes using our developed glass formulation study, the volume reduction factors for the IER, DAW and mixed wastes were evaluated as 21, 89 and 75, respectively.

• Korean Journal of Materials Research
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• v.9 no.11
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• pp.1137-1143
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• 1999

Microstructure and mechanical properties of Mod.9Cr-1Mo and W added 9Cr-0.5Mo2W steels were investigated for liquid metal reactor (LMR) heat exchange tube. The tempering temperatures at which cell structure was formed were $700^{\circ}C$ for Mod.9Cr-1Mo steel and $750^{\circ}C$ for W added 9Cr0.5Mo-2W steel. indicating the recovery of dislocation was delayed by the addition of W. 9Cr-0.5Mo-2W steel had the same kinds of precipitates with Mod.9Cr-1Mo steel, but the W was included in the precipitates in 9Cr-0.5Mo-2W steel. Micro-hardness and ultimate tensile strength of 9Cr-0.5Mo-2W steel were higher than those of Mod.9Cr-1Mo steel. The impact property of Mod.9Cr-1Mo steel was superior to that of 9Cr-0.5Mo-2W steel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.83-91
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• 2018

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria ($34kgf{\cdot}cm^{-2}$) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

• 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 $^{14}C$.

• 대한공업교육학회지
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• v.31 no.1
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• pp.200-210
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• 2006

This C-ring test, normally employed for evaluating susceptibility to stress-corrosion cracking, was determined to be a suitable small scale test to evaluate PWHT(Post-Weld Heat Treatment) cracking susceptibility. This test is possible to incorporate an actual weld, to introduce a notch into the coarse grained HAZ(Heat Affected Zone), to load the coarse grained HAZ any level of stress ad, most importantly, since the C-ring is an approximately constant strain type test, the stress decreases with time at temperature in a manner similar to that of an actual steel weldment. The procedure employed in making the C-ring was presented in the experimental procedure section, however, several points deserve further discussion. The walls of the weld groove are made along radial lines form the center of th var in order to obtain an HAZ which is oriented perpendicular to the walls of the machined C-ring. Therefore, the plane of maximum stress will be aligned through the HAZ and, therefore, crack propagation will not be forced to deviate form the plane of maximum stress in order to remain in the coarse grained HAZ as is the case with the Y groove test.

• Journal of Sensor Science and Technology
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• v.6 no.5
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• pp.385-391
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• 1997

An objection of this study is to develop a measuring circuit of a gauge using radioisotope for compaction control. The gauge developed in this study makes use of radioisotope with the activity exempted from domestic atomic law and consists of measuring circuits for gamma-rays and thermal neutrons, a high voltage supply unit, and a microprocessor. To obtain meaningful numbers of pulse counts, parallel five and two circuits are provided for gamma-rays and thermal neutrons, respectively. Being simple in electrical characteristics of G-M detector for gamma-rays, pulses are counted through only a shaping circuit. Very small pulses generated from He- 3 proportional detector for thermal neutrons are amplified to the maximum of 50 [dB] and a window comparator accepts only pulses with meaning. To minimize effects of natural environmental radiation and electrical noise, circuits are electrostatically shielded and pulses made by ripples are eliminated by taking frequency of high voltage supplied to the circuit and pulse height of ripples into consideration. One-chip microprocessor is applied to process various counts, results are stored and the gauage is made capable to communicate with PC. Enough and meaningful numbers of pulses are counted with the prototype gauage for compaction control.

• Korean Journal of Community Nutrition
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• v.13 no.2
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• pp.207-215
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• 2008

This study was conducted to measure and evaluate the food and nutrient intakes of the people living near the nuclear plant and to investigate the relationship between the household income level and the food and nutrient intake patterns. A total of 552 cases (263 males and 289 females) were surveyed during the period from April 1 to December 21 of 2005. Dietary intake was measured by means of the 24-hour recall method. The data were analyzed using SPSS Windows (ver. 14.0). The household income level of the subjects was classified into two groups : Low income group (LIG; $\leq$2,000,000 won) and high income group (HIG; > 2,000,000). The subjects at large had less energy and nutrient intakes than did the population in town and village who participated in the 2005 National Health and Nutrition Survey. The intake of calcium, zinc, vitamin A, riboflavin, vitamin $B_6$, vitamin C, and folic acid was less than the Estimated Average Requirement in case of $50{\sim}95%$ of the subjects. The LIG consumed less beans, vegetables, fruits, meats, and beverages than did the HIG in male, while the LIG consumed less eggs and beverages than did the HIG in female. The LIG consumed less nutrients than did the HIG in male, except for carbohydrate, while the LIG consumed less nutrients including zinc, vitamin A, riboflavin, vitamin B6, vitamin C, folic acid than did the HIG in female. In addition, the LIG had higher percentage energy consumption from carbohydrate. These results suggest that higher food and nutrient intake is associated with higher income.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.247-260
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• 2010

The developement of a HLW disposal canister is under way in KAERI using Cold Spray Coating technique. To estimate corrosion behavior of a cold sprayed copper, a creivice corrosion test was conducted at Southwest Research Institute(SWRI) in the United State. For the measurement of repassivation potential needed for crevice corrosion, three methods such as (1) ASTM G61-86 : Cyclic Potentiodynamic Polarization Measurements, (2) Potentiodynamic Polarization plus intermediate Potentiostatic Hold method, and (3) ASTM G192-08 (THE method) : Potentiodynamic- Galvanostatic-Potentiostatic Method, were introduced in this report. In the crevice corrosion test, the occurrence of corrosion at crevice area was optically determined and the repassivation potentials were checked for three kind of copper specimens in a simulated KURT underground water, using a crevice former dictated in ASTM G61-86. The applied electrochemical test techniques were cyclic polarization, potentiostatic polarization, and electrochemical impedance spectroscopy. As a result of crevice corrosion tests, every copper specimens including cold sprayed one did not show any corrosion figure on crevice areas. And the open-cell voltage, at which corrosion reaction initiates, was influenced by the purity of copper, but not their manufacturing method in this experiment. Therefore, it was convinced that there is no crevice corrosion for the cold sprayed copper in KURT underground environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.91-97
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• 2010

As nuclear power plants are getting older, interests on a decontaminating process are increasingly attracting more attention. Chemical decontamination is crucial to lower the production of radioactive waste and radiation dose rate. Prior to this, oxidizers and detergents for target material should be chosen so as to decontaminate major systems and components of a nuclear power plant chemically. In order to decontaminate it properly, it is crucial to have information about the chemical composition and crystalline structure of CRUD, analyzing its samples from the target or the decontamination system with components. However, there is no program which enables the extraction of samples directly from the object or the decontamination system with components carrying genuine radioactivity. Therefore, it is limited to samples from corrosion products carrying partial radioactivity as a resource. The composition of CRUD varies considerably depending on refueling cycle because it is closely related to the constituent of basic material. After settling a target, it is crucial to analyze and obtain analytical information about CRUD as a decontamination target. In this paper, various technologies for manufacturing simulated CRUD are introduced as alternatives to unattained samples. A metal oxide or metal hydroxide was used to synthesize simulated cruds having chemical compositions and crystalline stricture similar to the actual one by 12 different methods. CRUD 4(metal oxides in the autoclave vessel) and CRUD 10(metal oxides in a crucible after hydrazing pretreatment)were chosen as the best method for Type 1 and Type 2.respectively. As these CRUD can be synthesized easily without using any specialized equipment or reagents in a short time and in large quantities, they are expected to stimulate the development of decontaminating agents and processes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.113-119
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• 2011

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.