• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.255-261
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• 2022

A reliable and well-characterized dosimetry system which is traceable to the international measurement system, is the key element to quality assurance in radiation processing with cobalt-60 gamma rays, X-rays, and electron beam. This is specifically the case for health-regulated processes, such as the radiation sterilization of single use medical devices and food irradiation for preservation and disinfestation. Polyethylene is considered to possess a lot of interesting dosimetric characteristics. In this work, a detailed study has been performed to determine the dosimetric characteristics of a commercialized high-density polyethylene (HDPE) film using Fourier transformed infrared spectrometry (FTIR). Correlations have been established between the absorbed dose and radiation induced infrared absorption in polyethylene having a maximum at 965 cm^-1 (transvinylene band) and 1716 cm^-1 (ketone-carbonyl band). We have found that polyethylene dose-response is linear with dose for both bands up to1000 kGy. For transvinylene band, the dose-response is more sensitive if irradiations are made in helium. While, for ketone-carbonyl band, the dose-response is more sensitive when irradiations are carried out in air. The dose-rate effect has been found to be negligible when polyethylene samples are irradiated with electron beam high dose rates. The irradiated polyethylene is relatively stable for several weeks after irradiation.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.681-688
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• 2022

Dose monitoring in CT patients requires accurate dose estimation but most of the CT dose calculation tools are based on Caucasian computational phantoms. We established a library of organ dose conversion coefficients for Korean adults by using four Korean adult male and two female voxel phantoms combined with Monte Carlo simulation techniques. We calculated organ dose conversion coefficients for head, chest, abdomen and pelvis, and chest-abdomen-pelvis scans, and compared the results with the existing data calculated from Caucasian phantoms. We derived representative organ doses for Korean adults using Korean CT dose surveys combined with the dose conversion coefficients. The organ dose conversion coefficients from the Korean adult phantoms were slightly greater than those of the ICRP reference phantoms: up to 13% for the brain doses in head scans and up to 10% for the dose to the small intestine wall in abdominal scans. We derived Korean representative doses to major organs in head, chest, and AP scans using mean CTDI_vol values extracted from the Korean nationwide surveys conducted in 2008 and 2017. The Korean-specific organ dose conversion coefficients should be useful to readily estimate organ absorbed doses for Korean adult male and female patients undergoing CT scans.

• 방사성폐기물학회지
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• 제20권4호
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• pp.511-521
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• 2022

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y^-1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

• 방사성폐기물학회지
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• 제18권2_spc호
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• pp.317-325
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• 2020

한국원자력환경공단은 처분시설 내 1단계 인수·저장구역의 인수검사 공간 및 드럼 취급 공간 부족에 대한 문제를 해결하기 위하여 방폐물검사건물을 건설하여 저장·처리능력을 확충할 예정이다. 본 연구에서는 MCNP 코드를 이용하여 방폐물검사건물 내 저장구역에서 취급하는 해체 방사성폐기물 대상 신형처분용기를 대상으로 작업종사자의 피폭선량을 평가하였다. 평가결과, 시설 내 저장 가능한 최대 용기 개수(304개)와 방사선작업에 대한 연간 예상 작업시간(약 306시간)에 대하여 연간 집단선량은 총 84.8 man-mSv로 계산되었다. 시설 내 총 304개의 신형처분용기(소형/중형 타입)가 저장 완료된 시점에서 인수검사, 처분검사를 위한 작업종사자의 투입인력은 총 25명, 작업종사자 당 예상피폭선량은 연평균 3.39 mSv로 산출되었다. 소형용기 취급 시 작업종사자의 고방사선량 작업에 따른 작업효율과 방사선적 안전성 확보를 위해서는 콘크리트 라이너의 두께를 증가시키는 추가적인 차폐가 필요할 것으로 평가되었다. 향후 본 연구를 바탕으로 실측기반의 해체폐기물의 선원항과 특성을 활용하여 방사선작업 당 작업시간 및 투입인력을 산출함으로써 작업종사자의 최적의 방사선작업조건을 도출할 수 있을 것으로 사료된다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.339-347
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• 2005

IP-2형 운반용기는 정상운반조건의 낙하시험을 거친 후에 방사성내용물의 유실 또는 분산이 없고 외부표면에서의 방사선량률이 $20\%$ 이상 증가할 수 있는 차폐능력의 상실이 없어야 한다. 차폐체의 두께가 큰 경우에는, 볼트체결방식을 사용하면 IP형 운반용기로 많이 사용되고 있는 ISO 컨테이너에서의 문을 사용하는 경우보다 방사성내용물의 유실 또는 분산이 일어나지 않을 것이다. 본 연구에서는 볼트체결방식의 IP-2형 운반용기에 대한 낙하시험을 실시하기 전에, 낙하방향에 따른 볼트인장력과 볼트체결력을 시험으로 평가하는 방법을 알아보기 위하여 예비시험모델을 제작하여 시험 및 평가하였다. 바닥과 뚜껑방향의 수직낙하, 수평낙하, 4방향의 경사낙하 등 총 7가지 낙하방향으로 자유낙하를 실시하였다. 낙하충격에 의한 볼트인장력은 힘센서를 이용하여 측정하였고, 볼트체결력은 낙하시험 전의 체결토르크에서 낙하충격 이후의 풀림토르크의 변화를 이용하여 알아보았다.

• 방사성폐기물학회지
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• 제15권4호
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• pp.381-390
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• 2017

국내 원자력발전소의 사용후핵연료는 소내저장시설에 보관되어 있으나 저장시설의 용량 확장이 어려움이 있으며, 연구기관의 연구로에서 발생하는 고준위 방사성폐기물도 자체 보관중이나 영구적으로 저장할 수 없다. 또한 원전의 해체 시에도 고준위 방사성폐기물이 발생할 것으로 예상된다. 이에 따라, 본 연구에서는 현재 개발된 사용후핵연료 운반용기를 사용하여 고준위 방사성폐기물을 가상의 관리시설로 철도를 통하여 운반하는 경우에 대하여 작업자 및 운반경로 주변 일반인의 예상 피폭선량을 평가하였으며, 그 결과를 국내 법적기준치와 비교하였다. 또한, 고준위 방사성폐기물의 상하차 작업 시 작업자와 운반용기 간 거리와 운반사고 시 방사성핵종의 누출율의 변화에 따른 피폭선량의 변화에 따른 피폭선량 추이와 운반에 사용되는 열차의 구성에 따른 운반작업자의 피폭선량 변화를 분석하였다. 본 연구에서 설정한 모든 조건에서의 예상피폭선량은 국내 법적제한치 이하임을 확인하였다.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2016년도 제50회 동계 정기학술대회 초록집
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• pp.414.2-414.2
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• 2016

The optimal configuration of a transmutation reactor based on a low aspect ratio tokamak is determined using coupled analysis of tokamak systems and neutron transport. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on a neutron multiplication, a tritium-breeding ratio, and a power density. It is shown that a breeding blanket model has an impact on the radial build of a transmutation blanket. A burn cycle has to be determined to limit a fast neutron fluence of a plasma facing material below a radiation damage limit.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.151-152
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• 2004

An acceleration technique combined with the discrete ordinates method which has been widely used in the solution of neutron transport phenomena is applied to the solution of radiative transfer equation. The self-adjoint form of the second order radiation intensity equation is used to enhance the stability of the solution, and a new linearization method is developed to avoid the nonlinearity of the material temperature equation. This new acceleration method is applied to the well known Marshak wave problem, and the numerical result is compared with that of a non-accelerated calculation

• 한국연소학회:학술대회논문집
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• 한국연소학회 2012년도 제44회 KOSCO SYMPOSIUM 초록집
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• pp.97-98
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• 2012

The rotary kiln is one of the most widely used industrial reactors for contacting gases and solids. Kilns are mainly used for drying, calcining and reducing solid materials. In an indirected fired rotary kiln, heat is supplied to the outside of the kiln wall. Heat transfer in indirected fired rotary kilns encompasses all the modes of transport mechanisms, that is, conduction, convection and radiation. This paper deal with the heat transfer characteristics of indirect fired rotary kiln for scale up.