• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.57-65
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• 2005

This study suggested a new method for the solidification of molten salt waste generated from the electro-metallurgical process in the spent fuel treatment. Using binary material system, sodium silicate and phosphoric acid, metal chlorides were converted into metal phosphate in the micro-reaction module formed by SiO$_{2} particles. The volatile element in the reaction module would little vaporized below 1100$^{circ}$C After the gel product was mixed with borosilicate glass powder and thermally treated at 1000$^{circ}$C, li exists as Li$_{3}$PO$_4$ separated from glass phase and, Cs and Sr would be incorporated into an amorphous phase from XRD analysis. In case of the addition of ZrCl$_{4}$ to the binary system, the gel products were transformed into NZP structure considered as an prospective ceramic waste form after heat-treatment above 700 $^{circ}$C. From these results, the gel-route pretreatment can be considered as an effective approach to the solidincation of molten salt waste by the confirmed process or waste form and this also would be an alternative method on the ANL method using zeolites in USA by the confirmation of its chemical durability as an future work.

• Korean Chemical Engineering Research
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• v.51 no.6
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• pp.700-708
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• 2013

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. Pyroprocessing has been developed with the dry technologies which are performed under high temperature conditions excluding any aqueous processes. Pyro-processes which are based on the electrochemical principles require pretreatment processes and a voloxidation process is considered as a pretreatment step for an electrolytic reduction process. Various kinds of gas conditions are applicable to the voloxidation process and the understanding of Cs behavior during the process is of importance for the analyses of waste characteristics and heat load on the overall pyroprocessing. In this study, the changes of chemical compounds with the gas conditions were calculated by analyzing gas-solid reaction behavior based on the chemical equilibria on a Cs-Te-O system. $Cs_2TeO_3$ and $Cs_2TeO_4$ were selected after a Tpp diagram analysis and it was confirmed that they are relatively stable under oxidizing atmospheres while it was shown that Cs and Te would be removed by volatilization under reducing atmosphere at a high temperature. This work provided basic data for predicting Cs behavior during the voloxidation process at which compounds are chemically distributed as the first stage in the pyroprocessing and it is expected that the results would be used for setting up material balances and related purposes.

• Proceedings of the Korean Society for Agricultural Machinery Conference
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• 2002.07a
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• pp.195-201
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• 2002

입고ㆍ예냉ㆍ출고까지 전 공정을 자동화하고 진공압력을 작물의 품온에 따라 능동적으로 제어할 수 있고 진공챔버내에 콜드트랩을 설치하여 냉각효율을 향상시킨 진공예냉장치를 개발하여 고랭지 배추와 느타리버섯, 상추를 대상으로 예냉성능을 평가하고 예냉이 이들 작물에 미치는 영향을 분석하였다. 가. 새로 개발된 진공식 예냉장치는 공급컨베이어, 좌우 자동 슬라이딩 문, 이송컨베이어, 진공챔버, 진공펌프 콜드트랩, 냉동기로 구성되어 있다. 팔레트를 공급컨베이어에 올려놓고 작동을 시작하면, 입구문이 열리고 팔레트가 진공챔버내로 이송되면, 진공펌프에 의해 진공챔버내의 압력을 떨어뜨리고, 콜드트랩과 냉동기가 작동되어 예냉이 시작된다. 예냉이 완료되면 출구문이 열리고 이송컨베이어가 역회전하여 밖으로 배출되도록 되어있다. 나. 제작된 예냉장치의 예냉성능을 평가하기 위하여 느타리버섯, 고랭지 배추, 상추를 대상으로 냉각속도, 냉각균일도, 예냉후 저장중 품질변화시험을 실시하였다. 다. 시험결과 냉각소요시간은 느타리버섯의 경우 초기품온 15.2$^{\circ}C$에서 $1.5^{\circ}C$까지 냉각시키는데 24분, 고랭지배추는 13.5$^{\circ}C$에서 3.7$^{\circ}C$까지 냉각시키는데 18분, 상추는 13.4$^{\circ}C$에서 2.$0^{\circ}C$까지 냉각시키는데 24분 소요되었다. 평균냉각속도는 느타리버섯이 34.3$^{\circ}C$/h, 고랭지배추 32.6$^{\circ}C$/h, 상추 28.5$^{\circ}C$/h로 나타났다. 라. 또한 각층간의 냉각균일도를 알아보기 위하여 포장상자내에서 표면 품온과 내부품온변화를 조사한 결과 차이가 거의 없어 균일한 냉각이 가능하였다.생기 양단의 온도차는 높게 나타났고, 재생기 양단의 압력 차는 낮게 나타났다. 재생기 축열재로서 철망-철선을 사용할 경우 철선-철망 ø1.2-150이 전열 표면적은 작으나 재생기 양단의 온도차가 가장 큰 것으로 나타났으며 재생기 양단의 압력 차는 가장 낮게 나타나 공시 철망- 철선 혼합 축열재중 가장 우수함을 알 수 있다. 4. 철망사이에 철선을 삽입한 축열재의 경우, 철망사이에 삽입한 철선의 직경이 큰 것이 철선의 직경이 작은 것보다 재생기의 양단의 온도차가 높게 나타났고 재생기 양단의 압력차는 작게 나타났다. 그러므로 철망사이에 철선을 삽입한 것 중 성능이 우수한 것은 150-ø2. 0-150으로 나타났다. 5. 실험한 재생기 축열재들 중에서 성능이 우수한 것들을 비교한 결과, 복합 철선 ø1.2-1 50이 가장 성능이 좋은 것으로 나타났다.적외선.열풍 복합건조방법이 높게 나타나 이것은 곡물 표면에 원적외선 방사에의한 복사열이 전달되어 열장해를 받았기 때문으로 판단되며, 금후 더 연구하여 적정 열풍온도 및 방사체 크기를 구명해야 할 것이다.으로 보여진다 따라서 옻나무 유래 F는 포유동물의 생식기능에 중요하게 작용하는 것으로 사료된다.된다.정량 분석한 결과이다. 시편의 조성은 33.6 at% U, 66.4 at% O의 결과를 얻었다. 산화물 핵연료의 표면 관찰 및 정량 분석 시험시 시편 표면을 전도성 물질로 증착시키지 않고, Silver Paint 에 시편을 접착하는 방법으로도 만족한 시험 결과를 얻을 수 있었다.째, 회복기 중에 일어나는 입자들의 유입은 자기폭풍의 지속시간을 연장시키는 경향을 보이며 큰 자기폭풍일수록 현저했다. 주상에서 관측된 이러한 특성은 서브스톰 확장기 활동이 자기폭풍의 발달과 밀접한 관계가 있음을 시사한다.se that were all low in two aspects, named "the Nonsignificant group". And th

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.189-199
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• 2004

A bridge transported servo manipulator (BTSM) system for the advanced spent fuel conditioning process (ACP) has been developed to overcome the limitation of access, which is a drawback of mechanical master-slave manipulators (MSM) for the equipment maintenance. The servo manipulator is composed of a slave manipulator attached to the telescoping tubesets equipped with the overhead bridge installed at a hot cell and a master manipulator installed at an out-of-hot cell. Each manipulator has 7 degrees-of-freedom (DOF): a body rotation, an upper-arm tilt, a lower-arm tilt, a lower-arm rotation, a wrist pan & tilt, and a grasp motion. A wire-driven mechanism for a lower-arm rotation, a wrist pan and tilt, and a grasp motion of the manipulator has been adopted to increase the handling capacity compared to the manipulator weight and decrease the friction. The main disadvantage of the wire-driven mechanism is that if one link is in motion, other links can be affected. In this paper, the transmission characteristics among the wire-driven links have been formulated to overcome this drawback. The unexpected behaviors are confirmed by analyses of transmission characteristics as well as experiments. Also, the experimental results show that the unexpected behaviors are greatly decreased by the proposed compensation equations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.263-269
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• 2015

It is necessary to investigate the electrodeposition behavior of uranium and other elements on the cathode in the electrorefining process to recover the uranium selectively from the reduced metals of the electrolytic reduction process since transuranic elements and rare earth elements is dissolved in the LiCl-KCl eutectic salt. Study on separation factors of U, Ce, Y and Nd based on U and Ce was performed to investigate the deposition behavior of the cathode with respect to the concentration of rare earth elements in LiCl-KCl eutectic salt. After electrorefining with constant current mode by using Ce metal as a sacrifice anode, the contents of U, Ce, Y and Nd in the salt phase and the deposit phase of the cathode were analyzed, and separation factors of the elements were obtained from the analyses. Securing conditions of pure uranium recovery in the elctrorefining process was investigated by considering the separation factors with respect to $UCl_3$ and $CeCl_3/UCl_3$ ratio.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.423-434
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• 2016

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

• Journal of the Korean Geotechnical Society
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• v.33 no.10
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• pp.25-31
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• 2017

The buffer is one of the major components of an engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW). As the buffer is located between a disposal canister and host rock, it is indispensable to assure the disposal safety of high-level radioactive waste. It can restrain the release of radionuclide and protect the canister from the inflow of groundwater. Since high quantity of heat from a disposal canister is released to the surrounding buffer, thermal properties of the buffer are very important parameters for the analysis of the entire disposal safety. Especially, temperature criteria of the compacted bentonite buffer can affect the design of HLW repository facility. Therefore, this paper investigated thermal properties for the Kyungju compacted bentonite buffer which is the only bentonite produced in South Korea. Hot wire method and dual probe method were used to measure thermal conductivity and specific heat capacity of the compacted bentonite buffer according to the temperature variation. Thermal conductivity and specific heat capacity were decreased dramatically when temperature variation was between $22^{\circ}C{\sim}110^{\circ}C$ as degree of saturation decreased according to the temperature variation. However, there was little variation under the high temperature condition at $110^{\circ}C{\sim}150^{\circ}C$.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.115-120
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• 2009

Since 1990s, some events - detection of a dirty bomb in a Russian nation park in 1995, 9/11 terrorist attack to WTC in 2001, discovery of Al-Qaeda's experimentation to build a dirty bomb in 2003 etc - have showed that nuclear or radiological terrorism relating to radioactive materials (hereinafter "radioactive materials" is referred to as "nuclear material, nuclear spent fuel and radioactive source") is not incredible but serious and credible threat. Thus, to respond to the new threat, the international community has not only strengthened security and physical protection of radioactive materials but also established prevention of and response to illicit trafficking of radioactive materials. In this regard, our government has enacted or revised the national regulatory framework with a view to improving security of radioactive materials and joined the international convention or agreement to meet this international trend. For the purpose of prevention of nuclear/radiological terrorism, this paper reviews physical characteristics of nuclear material and existing detection instruments used for prevention of illicit trafficking. Finally, national detection regime against nuclear/radiological terrorism based on paths of the smuggled radioactive materials to terrorist's target building/area, national topography and road networks, and defence-in-depth concept is suggested in this paper. This study should contribute to protect people's health, safety and environment from nuclear/radiological terrorism.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.65-70
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• 2010

The distillation rate on LiCl-KCl eutectic salt under different vacuums from 0.5-50 mmHg was first investigated by using both a non-isothermal and a isothermal thermogravimetric (TG) analysis. Based on the non-isothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A distillation rate of $10^{-4}-10^{-5}$ mole $cm^{-2}sec^{-1}$ is obtainable at temperatures less than 1300K and vacuums of 0.5-50 mmHg. About a 99% salt distillation efficiency was obtained after an hour at a temperature above 1150 K under 50 mmHg in a small scale distillation test system. An increase in the vaporizing surface area is relatively effective for removing residual salt in the remaining particles, when compared to that for the vaporizing time. Over 99.95% of total distillation efficiency was obtained for a 1-h distillation operation by increasing the inner surface area from $4.52cm^2$ to $12.56cm^2$.