• Proceedings of the Korean Reliability Society Conference
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• 2006.05a
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• pp.239-246
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• 2006

프레스다이용 코일스프링은 자동차 및 전자제품의 외형생산에 필요한 금형 내에 장착되는 금형용 스프링으로서 녹아웃 및 스트리퍼 등에 사용되고 있다. 프레스다이용 코일스프링이 사용 중 파손 시에는 고가인 금형의 손상 및 생산성에 영향을 미칠 수도 있기 때문에 사용 환경에서의 신뢰성확보가 요구되어지고 있다. 특히 중(重)하중 및 극중(極重)하중용 스프링은 과거 현장에서 파손사례가 자주 발생함으로 인해 외산을 선호하는 경향이 있는 형편이다. 이에 국산 스프링의 신뢰성검증 및 확보를 위해 신뢰성기반구축사업을 통해 신뢰성평가기준(RS D 0014)가 제정되었으며, 이 평가기준에 의거하여 국내 업체의 제품에 대해 신뢰성평가를 실시하였다. 프레스다이용 코일스프링의 파손원인은 주로 반복하중에 의한 피로파손과 일정한 변위의 변형으로 발생하는 코일스프링 자유높이의 축소로 크게 구분되어질 수 있다. 시험결과 주 파손양상은 피로에 의한 균열발생이었으며, 코일 끝단부와 끝단부 직하부의 코일과의 마찰에 의한 균열발생이 주원인이었다. 즉, 코일의 끝단면과 직하면 코일이 연속적으로 부딪침으로써 발생한 변형 및 마모에 의해 표면균열이 발생하고, 표면균열에서 반복적인 부하하중이 가해짐으로써 피로균열 진전을 통해 점차적으로 파손이 진행되어졌음을 알 수 있었다. 본 발표에서는 기준에 의거하여 로하중용 프레스 다이용 코일스프링을 평가한 신뢰성평가시험 결과에 대해 보고하고, 파단면 관찰과 외산제품과의 미세조직 및 조성 등의 비교분석결과 등을 기초로 파손원인을 분석한 결과에 대해 보고하고자 한다.제고할 수 있을 것으로 기대한다.X>$CdCl^+,\;CdSO_4$ 등이 형성되었다. 수은의 경우는 해수 및 증류수를 용출용매로 이용한 모든 경우에서 납, 구리, 카드뮴과는 달리 대부분 침전하였다. 더욱이 해수에 존재하는 고농도 염소($Cl^-$)와의 수착으로 인해 finite solid인 calomel($Hg_2Cl_2$)이 형성되어 대부분 침전(SI=0)되기 때문에 납, 구리, 카드뮴 보다 더 낮은 환경이동성을 갖을 것으로 사료된다. 상기 실험결과 용출용매로 증류수와 해수를 이용했을 때, 제강 슬래그에서 용출되는 납, 구리, 카드뮴, 수은의 용출 경향의 차이를 확인할 수 있었고 이에 따라서, 납, 구리, 카드뮴의 용출 유해성은 낮기 때문에 해양구조물로의 제강슬래그 유효이용은 적합할 것으로 판단되었다.im80%$로 계산되었다. 열형광선량계로 측정된 방사선량은 각각 1.8, 1.2, 0.8, 1.2, 0.8 (70 cm 거리) cGy로 측정되었으며, 환자의 복부 표면에서의 서베이메터를 이용한 측정량은 10.9 mR/h였다. 차폐구조물의 사용 시 전체 치료 동안에 태아선량은 약 1 cGy 정도로 평가되었다. 결론 : AAPM Report No.50의 자료에 따르면, 임산부의 방사선 치료 시 태아의 방사선 피폭선량은 5 cGy 이하일 경우에 방사선 피폭에 따른 태아의 위험이 거의 없는 것으로 제시되고 있다. 본원에서 차폐 구조물을 설치하였을 경우에 측정된 태아선량은 약 1 cGy로 측정되었고, 고안된 차폐구조물

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.441-454
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• 2018

Since the permanent shutdown of Kori Unit 1 in 2017, a full-scale decommissioning project for a commercial nuclear reactor has been approaching. It is estimated that about 160,000 t of low-activity concrete waste will be produced from decommissioning of one unit of this commercial nuclear power reactor. Accordingly, it is necessary to review whether the effectiveness of the current regulatory framework for clearance waste (i.e. waste stream that meets activity concentration guidelines or dose criteria for clearance set forth in NSSC Notice No. 2017-65) can be maintained for the clearance of a bulk amount of concrete waste. In this regard, the IAEA SRS No. 44, which was used as a basis for revision of the Korean clearance regulations, is thoroughly analyzed and the radiological effects from four different clearance scenarios, along with input values and parameters derived from industrial practices in Korea, were evaluated. Though it is shown that the maximum annual dose from most recycling scenarios will be less than the clearance dose criterion for the normal scenario (i.e. an order of magnitude of $0.01mSv{\cdot}y^{-1}$), the radiation dose, estimated with conservative assumptions for the banking scenario, may exceed the above clearance dose criteria. Therefore, for safe and sustainable clearance of the bulk amount of concrete waste, it is required to diversify the concrete waste processors, perform more detailed site-specific assessment, and apply limiting conditions to the banking scenario.

• Journal of Radiation Protection and Research
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• v.1 no.1
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• pp.1-9
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• 1976

A suggestion has been made for neutron dosimetric techniques using activation and threshold detectors in criticality accidents. Neutron dosimetrical parameters, namely, the fission spectrum-averaged cross-sections of some threshold reactions and fluence-to-dose conversion factors have been calculated by the use of an electronic computer. It appears that detectors having comparatively high threshold energy give more fine information on spectral deformation in criticality accidents, while detectors with low threshold energy are of usefulness for measuring fast neutron fluence regardless of fissioning types. Unexpectedly it is found that the fission spectrum-averaged cross sections of the $^{32}S(n,\;p)^{32}P$ reaction is not sensitive to analytical forms of fission neutron spectrum: the modified Cran-berg and Maxwellian forms. In addition, the fluence-to-dose conversion factors seem to be insensitive to both spectral functions and fissioning types.

• Progress in Medical Physics
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• v.2 no.2
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• pp.155-160
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• 1991

We have studied the effective point of measurement for cylindrical ion chamber in water phantom for medical electron beams. Markus parallel plate chamber water phantom are used for the measurement of depth dose to determine the depth of the effective point of measurement for various energies(for electron 6MeV, 9MeV, 12MeV, 16MeV, and 20MeV; Co-60; for photon 6MV, 15MV). Cylindrical ion chambes(PTW233643 with r=2.75mm, PR-05P with r=2mm, and PM30 wiht r=15mm are used for the measurement of depth dose by same mtethod and the values of d$\_$50/ and R$\_$p/ obtained by three cylindrical chambers were compared with those of a flat chamber. From this we could evaluate the effective measuring points of cylindrical ion chamber. The effective point of measurement was estimated as 0.4～0.6r shifted toward surface from the center of the chamber for electron beam, 0.3～0.7r for $\^$60/Co X-ray.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.377-387
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• 2018

Korea's first commercial nuclear power plant, Kori Unit 1, was permanently shut down on June 18, 2017, after 40 years of successful operation. Kori Unit 1 plans to construct a waste treatment facility in the turbine building prior to commencement of dismantling in earnest. Various radioactive wastes are decontaminated, disassembled, cut and melted in the waste treatment facility and sent to the radioactive waste repository. The proportion of metal radioactive waste in dismantled waste is about 70%, of which large metal radioactive waste is mainly generated in the primary circuit and has high radioactivity, so radiation exposure must be managed during disassembly. In this study, the steam generators are selected as large metal radioactive waste, the exposure doses of the dismantling workers are calculated using RESRAD-RECYCLE code and the methods for reducing the exposure doses are suggested.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.213-229
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• 2005

Nuclides in radioactive wastes are assumed to be transported in the geosphere by groundwater and probably discharged into the biosphere. Quantitative evaluation of doses to human beings due to nuclide transport in the geosphere and through the various pathways in the biosphere is the final step of safety assessment of the radioactive waste repository. To calculate the flux to dose conversion factors (DCFs) for nuclides appearing at GBIs with their decay chains, a template ACBIO which is an AMBER case file based on mathematical model for the mass transfer coefficients between the compartments has been developed considering material balance among the compartments in biosphere and then implementing to AMBER, a general and flexible software tool that allows to build dynamic compartment models. An illustrative calculation with ACBIO is shown.

• Journal of the Korean Society of Radiology
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• v.9 no.6
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• pp.375-379
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• 2015

The purpose of this study is to determine the effective energy of a polyenegetic X-ray beam. The half value layer(HVL) of aluminum for 80 kVp X-ray beam was measured by using optically stimulated luminescent nanoDot dosimeters(OSLnDs). The linear attenuation coefficient(${\mu}$) was calculated using the measured HVL. And the mass attenuation coefficient(${\mu}/{\rho}$) was obtained by dividing the linear attenuation coefficient by the density(${\rho}$) of aluminum. The effective energy($E_{eff}$) of the obtained mass attenuation coefficient was determined using data of the X-ray mass attenuation coefficients for photon energies of aluminum given by National Institute of Standards and Technology(NIST). As a result, the HVL value is 2.262 mmAl. The ${\mu}$ value is $3.06cm^{-1}$. The ${\mu}/{\rho}$ value is $1.114cm^2/g$. And the $E_{eff}$ value was determined at 29.79 keV.

• Progress in Medical Physics
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• v.16 no.2
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• pp.53-61
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• 2005

The quality factors for cylindrical ionization chambers for kV X-rays were determined by Monte Carlo calculation and measurement. In this study, the X-rays of 60-300 kV beam (lSO-4037) installed in KFDA and specified in energy spectra and beam qualities, and the chambers of PTW N23333 and N30001 were investigated. In calculations, the $R_{\mu}\;and\;R_{Q,Q_{0}}$ in IAEA dosimetry protocols were determined from the air kerma and the cavity dose obtained by theoretical and Monte Carlo calculations. It is shown that the N30001 chamber has a flat response of $\pm1.7\%$ in $110\~300kV$ region, while the response range of two chambers were shown to $\pm3\~4\%$ in $80\~250kV$ region. From this work we have discussed dosimetry protocol for the kV X-rays and we have found that the estimation of energy dependency is more important to apply dosimetry protocol for kV X-rays.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Radiation Oncology Journal
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• v.15 no.1
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• pp.71-78
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• 1997

Purpose : To collect beam data for dynamic wedge fields using conventional measurement tools without the multi-detector system, such as the linear diode detectors or ionization chambers. Materials and Methods : The accelerator CL 2100 C/D has two photon energies of 6MV and 15MV with dynamic wedge an91es of 15o, 30o, 45o and 60o. Wedge transmission factors, percentage depth doses(PDD's) and dose Profiles were measured. The measurements for wedge transmission factors are performed for field sizes ranging from $4\times4cm^2\;to\;20\times20cm^2$ in 1-2cm steps. Various rectangular field sizes are also measured for each photon energy of 6MV and 15MV, with the combination of each dynamic wedge angle of 15o 30o. 45o and 60o. These factors are compared to the calculated wedge factors using STT(Segmented Treatment Table) value. PDD's are measured with the film and the chamber in water Phantom for fixed square field. Converting parameters for film data to chamber data could be obtained from this procedure. The PDD's for dynamic wedged fields could be obtained from film dosimetry by using the converting parameters without using ionization chamber. Dose profiles are obtained from interpolation and STT weighted superposition of data through selected asymmetric static field measurement using ionization chamber. Results : The measured values of wedge transmission factors show good agreement to the calculated values The wedge factors of rectangular fields for constant V-field were equal to those of square fields The differences between open fields' PDDs and those from dynamic fields are insignificant. Dose profiles from superposition method showed acceptable range of accuracy(maximum 2% error) when we compare to those from film dosimetry. Conclusion : The results from this superposition method showed that commissionning of dynamic wedge could be done with conventional dosimetric tools such as Point detector system and film dosimetry winthin maximum 2% error range of accuracy.