본 논문에서는 몬스터에게 붙잡히지 않고 키를 찾아서 연구소를 탈출하는 공포 방탈출 게임을 제안한다. 제 안하는 게임을 진행한30명의 게임 플레이 로그 데이터 분석 결과와 설문조사 결과를 바탕으로, 제안하는 게 임의 특징을 분석한 결과는 다음과 같다. 첫째, 제안하는 게임은 다양한 아이템, 액션, 탈출 경로를 제공한다. 제안하는 게임이 숨을 곳도 많고 다양한 상호작용을 제공한다고 설문에서4점 이상 주었다. 또한, map의 footprint를 분석한 결과, 플레이어는 다양한 경로를 통해 키를 찾아서 탈출하였다. 둘째, 제안하는 게임은 외 관으로 기능을 추론할 수 있는 직관적인 오브젝트를 제공한다. 따라서, 플레이어는 시각적 공간 및 게임 아 이템 용도를 쉽게 파악하여 조작할 수 있다. 설문조사 결과에서, 플레이어는 조작감 관련 항목의 점수를4점 이상을 주었다. 셋째, 제안하는 게임에서 플레이어는 아이템이 충분할 때보다는 부족할 때 더 몰입을 잘 한 다. 게임 플레이 로그 데이터 분석 결과와 설문조사 결과에 따르면, 플레이어는 아이템이 부족할 때 더 크게 공포를 느끼고 상황에 몰입하여, 더 적극적으로 행동하게 되고 더 민감하게 반응한다.

Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

Spent ion exchange resins have been generated during the operation of nuclear facilities. These resins include radioactive nuclides. It is needed to fabricate them into a stable form for final disposal. Cement solidification process is a useful method for the fabrication of them into a waste form for final disposal. In this study, proper conditions for the fabrication of them into a stable waste form were determined using the cement solidification process. In-drum waste forms were then produced at the conditions, where the stability of representative samples was evaluated for final disposal. The samples were satisfied to the Waste Acceptance Criteria for low and intermediate level radioactive waste disposal sites. This result can be utilized to derive optimal conditions for the fabrication of spent ion exchange resins into a final disposal form.

Ring Tensile Test (RTT) is mainly performed for comparing tensile strength and total strain between nuclear fuel cladding specimens under various initial conditions. Through RTT, the loaddisplacement (F-D) curve obtained from the uniaxial tensile test can also be obtained. However, the Young’s modulus estimated from the gradient of the straight portion is much lower than general value of materials. The reasons include tensile machine compliance, slack in the fixtures, or elastic deformation of the fixtures and the tooling. Another reason is that the bending of the test part in the ring is stretched with two pieces of tools. Although the absolute value of the Young’s modulus is smaller than the actual value, it is applicable to calculate the ratio of the Young’s moduli of different materials, that is, the relative value. The Young’s modulus, or slope of the linear section, varies slightly depending on which location data is used and how much data is included. In order to obtain a more accurate ratio of Young’s moduli between materials using the RTT results, a post-processing method for the ring tensile test results that can prevent such human errors is proposed as follows. First, the slope of the linear section is obtained using the displacement and load when the load increase is the largest and the displacement and load of the position that is 95% of the maximum load increase. To replace the section where the ring-shaped specimen is stretched at the beginning of the F-D curve, a straight line equal to the slope of the linear section is drawn to the displacement axis from the position of maximum load increase and moved to the origin to obtain the final F-D curve for a RTT. Lastly, the yield stress uses the stress at the point where the 0.2% offset straight line and the F-D curve meet as suggested in the ASTM E8/E8M-11 “Standard test methods for tensile testing of metallic materials”. RTT results post-processing method was coded using FORTRAN language so that it could be performed automatically. In addition, sensitivity analysis of the included data range on the Young’s modulus was performed by using the included data range as 90%, 85%, and 80% of the maximum load increase.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

Many countries have used nuclear power to generate electricity. Uranium-235, which is used as fuel in nuclear power plants, produces many fission products. Among them, iodine-129 is problematic due to its long half-life (1.57×107 years) and high diffusivity in the environment. If it is released into the environment without any treatment, it could have a major impact on humans and ecosystems. Therefore, it must be treated into a stable form through capture and solidification. Iodine can be captured in the form of AgI through silver-loaded zeolite filters in off-gas treatment processes. However, AgI could be decomposed in the reducing atmosphere of groundwater, so it must be converted into a stable form. In this study, Al2O3, Bi2O3, PbO, V2O5, MoO3, or WO3 were added to the iodine solidification matrix, AgI-Ag2O-TeO2 glass. The glass precursors were mixed to the appropriate composition and placed in an alumina crucible. After heat treatment at 800°C for 1 hour, the melt was quenched in a carbon crucible. The leaching behavior and thermal properties of the glass samples were evaluated. The PCT-A test for leaching evaluation showed that the normalized releases of all elements were below 2 g/m2, which satisfied the U.S. glass wasteform leaching regulations. Diffrential scanning calorimetry (DSC) was performed to evaluate the thermal properties of all glass samples. The addition of MoO3 or WO3 to the AgI-Ag2O-TeO2 glass increased the glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability. The similar relative electro-static filed values of MoO3, and WO3 which are approxibately three times that of the glass network former TeO2, could provide sufficient force to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M=V, Mo, W) links. The high electrostatic forces of Mo and W increased the glass network cohension and prevented the crystallization of the glass.

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

For the spent fuel modeling, the plastic model of the cladding used in FRAPCON uses the σ􀷥 = K􀟝̃􀯡 􁉂 􀰌􁈶 􀬵􀬴􀰷􀰯􁉃 􀯠 format. Strength coefficient (K), strain hardening exponent (n), strain rate sensitivity constant (m) are derived as the function of temperature. The coefficient m related to the strain rate shows dependence on the strain rate only in the α-β phase transition section, 1,172.5~1,255 K. But this is the analysis range of the FRAPTRAN code, which is an accident condition nuclear fuel behavior evaluation code. It does not apply to evaluate spent fuel. This coefficient in FRAPCON is used as a constant value (0.015) below 750 K (476.85°C), and at a temperature above 750 K, it is assumed that it is linearly proportional to the temperature without considering the strain rate dependence, also. In order to confirm the effect of strain rate, multiple test data performed under various conditions are required. However, since the strain rate dependence is not critical and test specimen limitation in the case of spent fuel, it is needed to replace with a new plastic model that does not include the strain rate term. In the new plastic model, the basic form of the Ramberg-Osgood equation (RO equation) is the same as ε􀷤 = 􀰙􀷥 􀮾 + 􀜭􀯥 􁉀􀰙􀷥 􀮾􁉁 􀯡􀳝. If the new variable α is defined as α = 􀜭􀯥􁈺􀟪􀯢/􀜧􁈻􀯡􀳝􀬿􀬵, this equation can be transformed into ε􀷤 = 􀰙􀷥 􀮾 + 􀟙 􀰙􀷥 􀮾 􁉀 􀰙􀷥 􀰙􀰬 􁉁 􀯡􀳝􀬿􀬵 . The procedure for expressing the stress-strain curve of the cladding with the RO equation is as follows. First, convert the engineering stress-strain into true stress-strain. Second, using a data analysis program such as EXCEL or ORIGIN, obtain the slope of the linear trend-line on the linear part and use it as the elastic modulus. Third, using the 0.2% offset method, find the yield point and the yield stress. Finally, using the solver function of EXCEL, find the optimal values of α and 􀝊􀯥 that minimize the sum of errors. The applicability of the suggested RO equation was evaluated using the results of the Zircaloy-4 plate room temperature tensile test performed by the KAERI and the Zircaloy cladding uniaxial tensile test results presented in the PNNL report. Through this, the RO equation was able to express the tensile test results within the uncertainty range of ±0.005. In this paper, the RO equation is suggested as a new plastic model with limited test data due to the test specimen limitation of spent fuel and its applicability is confirmed.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

This study evaluates the gastroprotective effect of cabbage extract with sulforaphane content of 5.19 mg/L and Smethylmethionine content of 469.28 μg/L. In vitro, the lipopolysaccharide (LPS)-treated group had an increased NO activity compared to the normal group, and the concentration of NO was reduced when the cabbage extract was treated in the dose manner. The level of IL-6 induced by LPS was dose-dependently reduced when the extract was treated. The cabbage extract concentration was orally administered in rats at 5.75 mg/kg, 11.5 mg/kg, and 23 mg/ kg, and the inhibitory effect on gastric damage by HCl-ethanol was observed. Histological analysis exhibited mucosal erosion in the gastritis model compared to the normal group, while the ameliorating effect of the generated erosion was observed in the cabbage-treated group. The histamine concentration was significantly increased in the gastritis-induced animal model, and the histamine concentration was decreased in the 23 mg/L-treated group of cabbage extract. In conclusion, the results of this study suggest that cabbage extract not only down-regulates cytokines in vitro, but is also directly involved in histamine secretion in an animal model of gastritis; therefore, cabbage extract can help inhibit gastrointestinal disorders by improving the protective barrier.