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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
Journal Basic Information
Journal DOI :
The Korean Radioactive Waste Society
Editor in Chief :
Volume & Issues
Volume 5, Issue 4 - Dec 2007
Volume 5, Issue 3 - Sep 2007
Volume 5, Issue 2 - Jun 2007
Volume 5, Issue 1 - Mar 2007
Selecting the target year
Separation and Recovery for the Analysis of Radioiodine in RI Wastes
Kang, Sang-Hoon ; Han, Sun-Ho ; Lee, Heung-N. ; Jee, Kwang-Yong ; Lee, In-Koo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 267~272
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
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,
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
and 100 mL of 9 M
, and then distilled after adding 10 mL of 30%
and 1 mL of 30%
. The condensed iodine by circulator was extracted into
, then back-extracted into the aqueous phase with 10 mL of 5%
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
at this analytical condition was 0.6 Bq/g.
Conceptual Geochemical Modelling of Long-term Hyperalkaline Groundwater and Rock Interaction
Choi, Byoung-Young ; Yoo, Si-Won ; Chang, Kwang-Soo ; Kim, Geon-Young ; Koh, Yong-Kwon ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 273~281
Hyperalkaline groundwater formed by groundwater-cement components and its reaction with bedrock in a nuclear waste repository were simulated by geochemical modeling. The result of groundwater-cement components reaction showed that the pH of water was 13.3 and the precipitated minerals were Brucite, Katoite, Calcium Silicate Hydrate(CSH1.1), Ettringite, Hematite, and Portlandite. The result of interaction between such minerals and groundwater sampled in Gyeongju area also showed that the pH of groundwater reached 12.4. Interaction between such hyperalkaline groundwater and granite was simulated by kinetic model during
years. This result showed that the final pH of groundwater reached 11.2 and the variation of pH was controlled by dissolution/precipitation of silicate and CSH minerals. Groundwater quality was also determined by dissolution/precipitation of silicate, CSH, oxide minerals. Our results show that geochemical modeling of long-term hyperalkaline groundwater and rock interaction can contribute to the safety assessment of engineered barrier by predicting geochemical condition in repository site.
Investigation of Pyroprocessing Concept and Its Applicability as an Alternative Technology for Conventional Fuel Cycle
Yoo, Jae-Hyung ; Lee, Byung-Jik ; Lee, Han-Soo ; Kim, Eung-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 283~295
The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of
-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.
A Feasibility Study on the Polymer Solidification of Evaporator Concentrated Wastes
Yang, Ho-Yeon ; Kim, Ju-Youl ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 297~308
The granulation equipment of concentrated wastes is manufactured for the polymer solidification of concentrated wastes. It uses liquid sodium silicate as a granulating agent for the granulating of dried powder containing boric acid. The granulating agent is sprayed in the form of droplet and mean size of dried granules is
. The new technology which has been used for the polymer solidification of spent resin in U.S. and certified by Nuclear Regulatory Commission (NRC) is successfully applied to concentrated wastes. This uses in-situ solidification process within drum without mechanical mixing. Maximum loading of waste can be achieved without increasing of waste volume. Polymer waste forms were evaluated with several test such as fire test, compressive strength test, leaching test, immersion test, irradiation test, and thermal cycling test according to standard test procedures.
Reliability Evaluation of ACP Component under a Radiation Environment
Lee, Hyo-Jik ; Yoon, Kwang-Ho ; Lim, Kwang-Mook ; Park, Byung-Suk ; Yoon, Ji-Sup ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 309~322
This study deals with the irradiation effects on some selected components which are being used in an Advanced Spent Fuel Conditioning Process (ACP). Irradiation test components have a higher priority from the aspect of their reliability because their degradation or failure is able to critically affect the performance of an ACP equipment. Components that we chose for the irradiation tests were the AC servo motor, potentiometer, thermocouples, accelerometer and CCD camera. ACP facility has a number of AC servo motors to move the joints of a manipulator and to operate process equipment. Potentiometers are used for a measurement of several joint angles in a manipulator. Thermocouples are used for a temperature measurement in an electrolytic reduction reactor, a vol-oxidation reactor and a molten salt transfer line. An accelerometer is installed in a slitting machine to forecast an incipient failure during a slitting process. A small CCD camera is used for an in-situ vision monitoring between ACP campaigns. We made use of a gamma-irradiation facility with cobalt-60 source for an irradiation test on the above components because gamma rays from among various radioactive rays are the most significant for electric, electronic and robotic components. Irradiation tests were carried out for enough long time for total doses to be over expected threshold values. Other components except the CCD camera showed a very high radiation hardening characteristic. Characteristic changes at different total doses were investigated and threshold values to warrant at least their performance without a deterioration were evaluated as a result of the irradiation tests.
Status of the International Cooperation Project, DECOVALEX for THM Coupling Analysis
Kwon, Sang-Ki ; Cho, Won-Jin ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 323~338
For the assessment of the performance and safety of a deep underground radioactive repository system, the thermal, hydraulic, mechanical, and chemical behaviors and their coupling should be studied. In order to analyze the THMC coupling behavior more effectively, which requires complex mathematical models and modelling techniques, DECOVALEX international cooperation project was launched in 1992. Since its beginning, four major stages of the project were successfully completed and THMC modelling techniques for various conditions could be developed. In this study, the current status and major achievements from the project were reviewed and possible benefits of the participation to the project were discussed.
A Status of Technology and Policy of Nuclear Spent Fuel Treatment in Advanced Nuclear Program Countries and Relevant Research Works in Korea
You, Gil-Sung ; Choung, Won-Myung ; Ku, Jeong-Hoe ; Cho, Il-Je ; Kook, Dong-Hak ; Kwon, Kie-Chan ; Lee, Won-Kyung ; Lee, Eun-Pyo ; Hong, Dong-Hee ; Yoon, Ji-Sup ; Park, Seong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 339~350
Status on the spent nuclear fuel management policy and R&D plan of the major countries is surveyed. Also the prospect of the future R&D plan is suggested. Recently so-called fuel cycle nations, which have the reprocess policy of the spent fuel, announced new spent fuel management policy based on the advanced fuel cycle technology. The policy is focused to transmute highly radioactive material and material having a very long half-life, and to recycle the Pu and U contained in the spent fuel. In this way the radio-foxily of the spent fuel as well as the amount of the high level waste to be eventually disposed can greatly be reduced. Most of countries selected the wet process as a primary option for the treatment of the spent fuel since the advanced wet process, which is based on the existing PUREX process, looks more feasible as compared with the dry process. The wet process, however, is much more sensitive in terms of proliferation-resistance compared with the dry process. The pure Pu can easily be obtained by simply modifying the process. On the other hand the pure Pu can not be extracted in the dry process based on the high temperature molten salt process such as a pyroprocess. Even though the pyroprocess technology is very premature, it has a great merit. Thus it is necessary for Korea to have a long term strategy for pursuing a spent fuel treatment technology with a proliferation resistance and a great merit for the GEN-IV fuel cycles. Pyroprocess is one of the best candidates to satisfy these purposes.
A Study on Japanese Experience to Secure the Interim Storage Facility for Nuclear Spent Fuel
Kim, Kyung-Min ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 5, issue 4, 2007, Pages 351~357
The Japanese Government selected Mutsu, Aomori Prefecture as a provisional spent-fuel repository site. This comes as a result of the prefecture's five-year campaign to host the site since 2000. Korea stores spent nuclear fuel within sites of nuclear power plants, and expects the storage capacity to reach its limit by the year 2016. This compels Korea to learn the cases of Japan. Having successfully hosted Gyeongju as a site for low-to-intermediate-level nuclear waste repository, Korea has already learned the potential process of hosting spent fuel storage site. The striking difference between the two countries in the process of hosting the site is that the Korean government had to offer the local city a large amount of subsidy for hosting through competitive citizens' referendum among candidate cities while it was the leadership of the local municipality that enabled the controversial decision in Japan. It is also a distinguishable characteristics of Japan that not a huge subsidy is provided to the local host city. I hope this study offers an idea to Korea's future effort to select a spent-fuel host site.