• Journal of Radiation Industry
• /
• v.18 no.1
• /
• pp.89-93
• /
• 2024

Accidents at nuclear facilities and nuclear power plants led to leaks of large amounts of radioactive substances. Of the various radioactive nuclides released, ¹³⁷Cs are radioactive substances generated during the fission of uranium. Therefore, due to the high fission yield (6.09%), strong gamma rays, and a relatively long half-life (30 years), a rapid and efficient removal method and a study of adsorbents are needed. Accordingly, an adsorbent was prepared using Prussian blue (PB), a material that selectively adsorbs radioactive cesium. As a result of evaluating the adsorption performance with the prepared adsorbent, it was confirmed that 82% of the removal efficiency was obtained, and most of the cesium was rapidly adsorbed within 10 to 15 minutes. The purpose of this study was to adsorb cesium using the Prussian blue alginate bead and to compare the change in detection efficiency according to the amount of adsorbent added for quantitative evaluation. However, in this case, it is difficult to determine the detection efficiency using a standard source with the same conditions as the measurement sample, so the efficiency change of the HPGe detector according to the different heights of Prussian blue was calculated through MCNP simulation using certified standard materials (1 L, Marinelli beaker) for radioactivity measurement. It is expected to derive a relational equation that can calculate detection efficiency through an efficiency curve according to the volume of Prussian blue, quantitatively evaluate the activity at the same time as the adsorption of radioactive nuclides in actual contaminated water and use it in the field of nuclear facility operation and dismantling in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.4
• /
• pp.381-390
• /
• 2017

In Korea, commercial nuclear power plants and research reactors have on-site storage systems for the spent nuclear fuel, but it is difficult to expand the facilities used for the storage systems. If decommissioning of nuclear power plants starts, an amount of high level radioactive waste will be generated. In this study, a radiological impact assessment of the railroad transport of high level radioactive waste was carried out considering radiation workers and the public, using the developed transport container as the transport package. The dose rates for workers and the public during the transport period were estimated, considering anticipated transport scenarios, and the results compared with the regulatory limit. A sensitivity analysis was also carried out by considering the different release ratios of the radioactive materials in the high level radioactive waste, and different distances between the transport container and workers during loading and unloading phases and while attaching another freight car. For all the anticipated transport scenarios, the radiological impacts for workers and the public met the regulatory limits.

• Journal of Radiation Protection and Research
• /
• v.38 no.1
• /
• pp.1-9
• /
• 2013

Education and training is an important means of promoting safety culture and enhancing the level of competence of radiation worker in radiation protection. The existing international nuclear education and training of short duration has been carried out on the high-level officials and focussed on the classroom based training. The developing countries has been asking for support to cultivate their own technical experts to Korea which is a donor country exporting nuclear power plants. This paper summarizes the results of developing and operating the international education and training course to froster technical experts in radiation protection that emphasized practical training sessions and technical visits using the excellent domestic radiation facilities and infrastructure of education and training. It mentions the procedures of assessment and feedback as well. In an effort to maximize teaching-learning effects and to maintain consistency of the learning objectives, methods and assessment, SAT methodology has been applied on the processes of developing and operating the course. In the comparative and final assessment which were conducted at the beginning or at the end of training course, participants' average score increased around 2 points. The questionnaire of participants showed a high level of satisfaction of 4.0 points or above for the most of the questions. These imply teaching-learning methods applied to it might be effective. The teaching-learning methodologies may provide the opportunity to develop the customized training course for bringing up international technical experts and to shift educational paradigm from theory-oriented to on-site practice-based education.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.3
• /
• pp.239-245
• /
• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.268-279
• /
• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Journal of the Korean Society of Radiology
• /
• v.17 no.7
• /
• pp.1189-1195
• /
• 2023

In scenarios involving inspections and verifications of nuclear facilities, ensuring the proper functioning of on-site safeguards equipment is crucial. There have been precedents in Kazakhstan where equipment failed to operate properly due to extremly cold temperatures, and the year-round minimum temperature at North Korea's Punggye-ri nuclear test site is approximately minus 30 degrees Celsius. To ensure the proper functioning of equipment in extreme environments for on-site verification of nuclear activities on the Korean Peninsula, relevant research is necessary. This includes confirming the functionality of equipment used in inspections and verifications, as well as analyzing factors that may disrupt their normal operation. This study aims to conduct a risk analysis for the normal operation of equipment in extreme environments and develop criteria and procedures for environmental-based performance testing. To achieve this, we conducted a risk analysis based on IAEA safeguards, analyzed the utilization of equipment, and performed a risk analysis associated with transportation for on-site verification considering the environmental characteristics of the Korean Peninsula. Furthermore, we provided performance testing criteria and procedures. The research results can be utilized as reference material in the verification and monitoring processes of nuclear activities.

• Proceedings of the Earthquake Engineering Society of Korea Conference
• /
• 2000.04a
• /
• pp.226-232
• /
• 2000

Seismic probabilistic risk assessment(RA) rather than deterministic assessment provides more valuable information and insight for resolving seismic safety issues in nuclear power plant design. In the course of seismic PRA seismic fragility analysis is the most significant and essential phase especially for structural or mechanical engineers. Lately the seismic fragility analysis is taken as a useful tool in general structural engineering as well. A systemized and synthesized procedure or technology related to seismic fragility analysis of critical industrial facilities reflecting the unique experiences and database in Korea is urgently required. This paper gives a state-of-the-art reviews of PRA and briefly summarizes the technologies related to PRA and seismic fragility analysis before developing an unique technology considering characteristics of Korean database. Some key items to be resolved theoretically or technically are extracted and presented for the future research.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.10 no.4
• /
• pp.398-410
• /
• 1998

Two-phase flow phenomena are observed in many industrial facilities and make much importance of optimum design for nuclear power plant and the liquid transportation system. The particular flow pattern depends on the conditions of pressure, flow velocity, and channel geometry. However, the research on drag reduction in two-phase flow is not intensively investigated. Therefore, experimental investigations have been carried out to analyze the drag reduction and void fraction by polymer addition in the two-phase flow system. We find that the polymer solution changes the characteristic of two-phase flow. The peak position of local void friction moves from tile wall of the pipe to the center of the pipe when polymer concentration increase. And then we predict that it is closely related with the frau reduction.

• IE interfaces
• /
• v.23 no.3
• /
• pp.213-220
• /
• 2010

As the market becomes increasingly globalized and competition among companies increases in severity, various specialized organizations are participating across the process plant lifecycle, including the stages of design, construction, operation and maintenance, and dismantlement, in order to ensure efficiency and elevate competitiveness. In this regard, it is an important technical issue to develop services or information systems for sharing process plant data among participating organizations. ISO 15926 is an international standard for integration of lifecycle data for process plants including oil and gas facilities. ISO 15926 Part 7, a part of the ISO 15926 standard, specifies an implementation method called a facade that uses Web Services and ontology technologies for constructing plant data repositories and related services, with the aim of sharing lifecycle data of process plants. This paper discusses the ISO 15926-based prototype facade implemented for storing equipment data of nuclear power plants and servicing the data to interested organizations.

• Journal of Korean Institute of Industrial Engineers
• /
• v.28 no.2
• /
• pp.178-186
• /
• 2002

The objective of this study is to evaluate impact of computer-based man-machine interfaces of Korea Next Generation Reactor (KNGR) on the operator mental workload. Empirical experiments were conducted to measure the operator mental workloads of KNGR and Yong-Gwang Unit 3 and 4, respectively. A comparison analysis based on a NASA TLX revealed that Yong-Gwang Unit 3 and 4 were superior to KNGR in terms of the mental workload. Post-hoc analyses showed that the mental workload of senior reactor operators was significantly higher than those of reactor and turbine operators, regardless of plant types. The implications of the findings were discussed in detail.