• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.113-124
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• 2004

A computer software program, called BIDAS (BIoassay Data Analysis Software) is developed to interpret the bioassay measurement data in terms of intakes and the committed effective dose using the human respiratory tract model (HRTM), gastrointestinal tract (GI-tract) model and biokinetic models currently recommended by the International Commission on Radiological Protection (ICRP) to describe the behavior of the radioactive materials within the body. The program consists of three modules; first, a database module to manage the bioassay data, second, another databasee module to store the predicted bioassay quantities of each radionuclide and finally, a computational module to estimate the intake and committed effective dose calculated with the bioassay quantity measurement values from either an acute or chronic exposure of the radionuclies within the body. This paper describes the features of the program as well as the quality assurance check results of the BIDAS software program.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.9-16
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• 2009

Transport package for radioactive material should be structurally safe under puncture drop condition and its safety should be verified by test and numerical analysis. Most finite element analyses for puncture drop have been performed without modeling the impact limiter since failure is occurred in the materials of the impact limiter. This paper presents a new modeling methodology, where an element is eroded in case that the material's failure criteria are reached at the element's integration point, to investigate the effect of the impact limiter in the puncture process. The effectiveness of the proposed scheme is shown through the puncture drop analysis of hotcell transport cask, which is under design in KAERI. The results show that about 80 percent of the total impact energy is absorbed due to the deformation of impact limiter. Using the present method the puncture drop can be analyzed more accurately, but it would give conservative results compared to the actual test condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.349-358
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• 2005

In this study, analysis of the disposal tunnel spacing and disposal pit pitch was carried out, as a factor of the design to estimate the scale and layout of the repository. To do this, based on the reference repository concept and the engineered barrier concept, several cross sections of the disposal tunnel and disposal pit were established. After then, the mechanical and thermal stabilities of the established tunnels were analyzed. Also, an optimized disposal tunnel spacing and the disposal pit pitch reducing the excavation volume was proposed. The results of these analyses can be used in the deep geological repository design. The detailed analyses by the exact site characteristics data to reduce the uncertainty of the site and the modification for the optimization are required.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.281-294
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• 2012

The current status of the computer codes for the analysis of coupled thermal-hydrological-mechanical behavior occurred in a high-level waste repository was investigated. Based on the reported results on the comparison between the predictions using the computer codes and the experimental data from the in-situ tests, the reliability of the existing computer codes was analyzed. The presented codes simulated considerably well the coupled thermal-hydrological-mechanical behavior in the near-field rock of the repository without buffer, but the predictions for the engineered barrier system of the repository located at saturated hard rock were not satisfactory. To apply the current thermal-hydrological-mechanical models to the assessment of the performance of engineered barrier system, a major improvement on the mathematical models which analyze the distribution of water content and total pressure in the buffer is required.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.267-272
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• 2007

Various kinds of RI wastes are discharged from licensed organizations of radioisotopes les such as hospitals and clinic organizations, educational organizations, research institutions, and public organizations. Radioiodines such as $^{125}I\;and\;^{131}I$ are radioisotopes mainly used in nuclear medicine and industry. A method for the determination of radioiodines in RI wastes has been applied to measure low level activity using acid decomposition method and HPGe gamma ray spectrometer. Prior to analysis of real samples, $^{131}I$ reference solution and 10 g of yellow tissue paper was added to flask in mantle and was heated in 100 mL of 0.4 N $K_2Cr_2O_7$ and 100 mL of 9 M $H_2SO_4$, and then distilled after adding 10 mL of 30% $H_2PO_3$ and 1 mL of 30% $H_2O_2$. The condensed iodine by circulator was extracted into $CCl_4$, then back-extracted into the aqueous phase with 10 mL of 5% $K_2SO_2$ solution. Finally, $^{131}I$ was measured at 364.48 keV using HPGe gamma ray spectrometer after precipitation and filtration. Chemical yield of three steps such as acid decomposition process, chemical separation process, and precipitation and filtration process was more han 94% respectively, MDA(Minimum Detectable Activity) of $^{131}I$ at this analytical condition was 0.6 Bq/g.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.175-182
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• 2009

The spent cationic exchange resins and anionic exchange resins were separated from mixed spent exchange resins by a fluidized bed gravimetric separator. The separated resins were identified by an elemental analysis and thermogravimetric analysis. The each test sample was prepared by diluting the slurry made by wet ball milling the cationic exchange resins and the anionic exchange resins separated as a spherical granular form for 24 hours. The resulting test samples showed a slurry form of less than $75{\mu}m$ of particle size and 25,000ppm of $COD_{cr}$. The decomposition conditions of each test samples from a thermal power plant were obtained with a lab-scale(reactor volume : 220mL) supercritical water oxidation(SCWO) facility. Then pilot plant(reactor volume : 24 L) tests were performed with the test samples from a thermal power plant and a nuclear power plant successively. Based on the optimal decomposition conditions and the operation experiences by lab-scale facility and the pilot plant, a commercial plant(capacity : 150kg/h) can be installed in a nuclear power plant was designed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.231-237
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• 2017

The present study researched and analyzed decontamination technology for decommissioning a nuclear power plant. The decision-making technique (EXPERT-CHOICE) was used to evaluate and select the optimal decontamination technology. In principle, this evaluation method is generally performed by a group of experts in the relevant field. The results of the weights were calculated by multiplying the weights with regard to each criterion and evaluation score. The evaluation scores were categorized into 3 ranges (high, medium, and low), and each range was weighted for differentiation. The level of the technology analysis was improved by additionally quantifying the weights with regard to each criterion and subdividing criteria into subcriteria. The basic assumption of the evaluation was that the weight values would decided on in an expert survey and assigned to each criterion. The evaluation criteria followed high weight for the 'High' range. Accordingly, H, M, and L were assigned weights of 10:5:1, respectively. This was based on the EXPERT-CHOICE optimal analysis. The minimum and maximum values were excluded, and the average value was used as the evaluation value for each scenario.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.49-62
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• 2011

Recently, International Atomic Energy Agency and major leading countries in radioactive waste management tend to subdivide the categories of radioactive waste based upon risk-graded approach. In this context, the category of very low level waste has been newly introduced, or optimized management options for this kind of waste have been pursued in many countries. The application of engineered surface landfill type facilities dedicated to dispose of very low level waste has been gradually expanded, and it was analyzed that their design concept of isolation has been much advanced than those of the old fashioned surface trench-type disposal facilities for low and intermediate level waste, which were usually constructed in 1960's. In addition, the management options for very low level waste in major leading countries are varied depending upon and interfaced with the affecting factors such as: national framework for clearance, legal and practical availability of low and intermediate level waste repository and/or non-nuclear waste landfill, public acceptance toward alternative waste management options, and so forth. In this regard, it was concluded that optimized long-term management options for very low level waste in Korea should be also established in a timely manner through comprehensive review and discussions, in preparation of decommissioning of large nuclear facilities in the future, and be implemented in a systematic manner under the framework of national policy and management plan for radioactive waste management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.97-105
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• 2014

In the case of the direct measurement of $^{226}Ra$ using a HPGe gamma-ray spectrometer, the interference between gammarays with 186.21 keV of $^{226}Ra$ and 185.7 keV of $^{235}U$ should be corrected to calculate the net peak area in the energy spectrum. In general, it is very difficult to conduct peaks stripping with difference of about 0.5 keV, although a HPGe with the superior resolution is applied and the maximum channels is applied to the spectrometer. In this study, several interference correction techniques in the direct measurement were surveyed to evaluate the feasibility for the measurement of $^{226}Ra$ using the gamma-ray spectrometery. Applying the interference corrections to the analysis of raw materials and by-products, the method validation for the direct measurement of $^{226}Ra$ was conducted by evaluating the measurement uncertainty, linearity, and range. As a result, the optimum method of the interference correction was selected by comparing with the indirect measurement of which progenies of $^{226}Ra$, such as $^{214}Pb$ and $^{214}Bi$, were analyzed in the secular equilibrium state.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.125-133
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• 2017

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidification host matrix for the noble metal fission product waste in anode sludge from the electro-refining process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confirm that Zr-Cr based compositions will likely serve as chemically stable waste forms.