• Title/Summary/Keyword: Nuclear Fuels

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Factors Affecting the Minimum Detectable Activity of Radioactive Noble Gases (방사성 노블가스 측정을 위한 최소검출방사능 산출의 조절인자)

  • Park, Ji-young;Ko, Young Gun;Kim, Hyuncheol;Lim, Jong-Myoung;Lee, Wanno
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.16 no.3
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    • pp.301-308
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    • 2018
  • Anthropogenic radioactive noble gases formed by nuclear fission are significant indicators used to monitor the nuclear activity of neighboring countries. In particular, radioactive xenon, owing to its abundant generation and short half-life, can be used to detect nuclear testing, and radioactive krypton has been used as a tracer to monitor the reprocessing of nuclear fuels. Released radioactive noble gases are in the atmosphere at infinitesimal amounts due to their dilution in the air and their short half-life decay. Therefore, to obtain reliable and significant data when performing measurement of noble gases in the atmosphere, the minimum detectable activity (MDA) for noble gases should be defined as low as possible. In this study, the MDA values for radioactive xenon and krypton were theoretically obtained based on the BfS-IAR system by collecting both noble gases simultaneously. In addition, various MDA methods, confidence level and analysis conditions were suggested to reduce and optimize MDA with an assessment of the factors affecting MDA. The current investigation indicated that maximizing the pretreatment efficiency and performance maintenance of the counter were the most important aspects for Xe. In the case of Kr, since sample activities are much higher than those of Xe, it is possible to change the target MDA or to simplification of the analysis system.

Development of Induction Brazing System for Sealing Instrumentation Feedthrough Part of Nuclear Fuel Test Rig (핵연료조사리그 계장선 통과부위의 밀봉을 위한 유도 브레이징 시스템 개발)

  • Hong, Jintae;Kim, Ka-Hye;Heo, Sung-Ho;Ahn, Sung-Ho;Joung, Chang-Young;Son, Kwang-Jae;Jung, Yang-Il
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.37 no.12
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    • pp.1573-1579
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    • 2013
  • To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and $300^{\circ}C$ respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.

Domestic Greenhouse Gas Reduction Policy (국내 온실가스 감축 정책)

  • Bae, Sung-Ho
    • Journal of Energy Engineering
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    • v.20 no.1
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    • pp.8-12
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    • 2011
  • For reducing greenhouse gas emissions, the short-term strategy is of existing energy-efficient appliances to facilitate the spread of energy efficiency improvements to improve energy efficiency, energy saving projects that will include investments to enable. R&D is at the core of the long-term strategy. To reduce energy demand, the equipments and processes improved energy efficiency should be developed. In terms of energy supply, the policies for greenhouse gas reduction is to replace fossil fuels by expanding the supply of renewable energy such as solar, wind, geothermal, biomass and nuclear power as nearly zero-emission of greenhouse gas. In terms of energy consumption, measures to reduce greenhouse gas emissions is in line with the policy for efficiency improvement. The buildings & work-site of high-energy consumption in the building & Industry sectors, should implement a policy to strengthening the voluntary agreement on energy-saving facilities and expand to invest in energy saving facilities.

Study on Governing Equations for Modeling Electrolytic Reduction Cell (전해환원 셀 모델링을 위한 지배 방정식 연구)

  • Kim, Ki-Sub;Park, Byung Heung
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.12 no.3
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    • pp.245-251
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    • 2014
  • Pyroprocess for treating spent nuclear fuels has been developed based on electrochemical principles. Process simulation is one of the important methods for process development and experimental data analysis and it is also a necessary approach for pyroprocessing. To date, process simulation of pyroprocessing has been focused on electrorefining and there have been not so many investigations on electrolytic reduction. Electrolytic reduction, unlike electrorefining, includes specific features of gas evolution and porous electrode and, thus, different equations should be considered for developing a model for the process. This study summarized required concepts and equations for electrolytic reduction model development from thermodynamic, mass transport, and reaction kinetics theories which are necessitated for analyzing an electrochemical cell. An electrolytic reduction cell was divided and equations for each section were listed and, then, boundary conditions for connecting the sections were indicated. It is expected that those equations would be used as a basis to develop a simulation model for the future and applied to determine parameters associated with experimental data.

VARIATION OF NEUTRON MODERATING POWER ON HDPE BY GAMMA RADIATION

  • Park, Kwang-June;Ju, June-Sik;Kang, Hee-Young;Shin, Hee-Sung;Kim, Ho-Dong
    • Journal of Radiation Protection and Research
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    • v.34 no.1
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    • pp.9-14
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    • 2009
  • High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

The Construction Work Completion of the Fuel Test Loop (핵연료 노내조사시험설비 설치공사 완료)

  • Park, Kook-Nam;Lee, Chung-Young;Chi, Dae-Young;Park, Su-Ki;Shim, Bong-Sik;Ahn, Sung-Ho;Kim, Hark-Rho;Lee, Jong-Min
    • Proceedings of the KSME Conference
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    • 2007.05a
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    • pp.291-295
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    • 2007
  • FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL consists of In-Pile Test Section (IPS) and Out-Pile System (OPS). FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. Task Force Team was organized to do a construction systematically and the communication between members of the task force team was done through the CoP(community of Practice) notice board provided by the Institute. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. Without a sweet of the workers of the participating company of HEC(Hyundae Engineering Co, Ltd), HDEC(HyunDai Engineering & Construction Co. Ltd), equipment manufacturer, and the task force team, it is not possible to install the FTL facility within the planned shutdown period. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

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A Deterministic Safety Assessment of a Pyro-processed Waste Repository (A-KRS 처분 시스템 결정론적 안전성 평가)

  • Lee, Youn-Myoung;Jeong, Jongtae;Choi, Jongwon
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.10 no.3
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    • pp.171-188
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    • 2012
  • A GoldSim template program for a safety assessment of a hybrid-typed repository system, called "A-KRS," in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

Recovery of Residual LiCl-KCl Eutectic Salts in Radioactive Rare Earth Precipitates (방사성 희토류 침전물내 잔류하는 LiCl-KCl 공융염의 회수)

  • Eun, Hee-Chul;Yang, Hee-Chul;Kim, In-Tae;Lee, Han-Soo;Cho, Yung-Zun
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.8 no.4
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    • pp.303-309
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    • 2010
  • For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily.

Study On the Characteristics of Milled $UO_2$ Powder Prepared by Oxidation and Reduction Process (산화ㆍ환원처리된 $UO_2$ 분말의 분쇄특성 연구)

  • Lee Jae-Won;Lee Jung-Won
    • Resources Recycling
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    • v.11 no.4
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    • pp.3-10
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    • 2002
  • The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.

A Probabilistic Safety Assessment of a Pyro-processed Waste Repository (A-KRS 처분 시스템 확률론적 안전성 평가)

  • Lee, Youn-Myoung;Jeong, Jongtae
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.10 no.4
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    • pp.263-272
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    • 2012
  • A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.