• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.27-37
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• 1986

The neutron fluxes and dose rates due to radiation streaming from reactor cavities were evaluated at the KNU-1 reactor pressure vessel (RPY) head flange elevation. To find a suitable cross section data set for the evaluation, a benchmark test was performed for three data sets; DLC-23/CASK, DLC-31/FEWG, and DLC-47/BUGLE. The leakage fluxes from the KNU-1 RPV outer surface were calculated with two different methods: 1-D calculation with ANISN, and 2-D calculation with DOT3.5. The Monte Carlo procedures as embodied in the MORSE-CG code combined with the albedo option were applied to predict the radiation distributions in the cavity region. Finally, the activation analysis of the stud bolts was performed to identify the major activation products.

• Journal of Radiation Protection and Research
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• v.31 no.1
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• pp.37-46
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• 2006

For the first step to develop a reference disposal container of spent fuel to be used in a deep geological repository, this paper examined safe dimensions of the disposal container on the points of nuclear criticality and radiation safety and mechanical structural safety and provided basic information for dimensioning the container and configuration of the container components, and establishing the favorable and safe disposal conditions. When the safety factor for stress due to the external loads (hydrostatic and swelling pressure) is taken as 2.0, the safe diameter of the filler material to provide enough container strength under the assumed external loads is found to be 112cm with 13cm spacing between inner baskets in PWR container. Also the thickness of the thinner section between the fuel basket and the surface of the cast insert is determined to be 150 mm. Regarding these dimensions of the container, the PWR fuel container is sketched to accommodate 4 square assemblies or 297 CANDU fuel 297 bundles (33 circle tubes x 9 stacks). However the top and bottom parts need to be checked again through the detail radiation shielding analysis with respects to the emplacement position and handling processes of the disposal container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.37-43
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• 2012

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

• Proceedings of the Korea Concrete Institute Conference
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• 2003.05a
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• pp.985-990
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• 2003

In order to demonstrate Advanced Spent Fuel Conditioning Process (ACP), shielding facilities such as hot cell suitable to handling radionuclides and process property will be necessary. But the construction of new facilities needs much money, man-power and time, it is now scheduled to remodel the hot cell, which has already been installed and maintained at Irradiated Material Experiment Facility (IMEF) in the Korea Atomic Energy Research Institute (KAERI). The basic structure and concrete shield wall of hot cell partly have been constructed on the base floor in IMEF building in current status. And hot cell after remodeling will be used for carrying out the lab-scale experiment of ACP. The hot cell was built in accordance with 35 curies of fe-59(1.2 MeV) as design criteria of radiation dose limit. But the radioactive source of ACP is expected to be much higher than design criteria of IMEF, shielding ability of the hot cell in the current status is unsatisfactory to the hot test of ACP. Therefore shield wall shall be reinforced with heavy concrete, steel or lead. In this paper, dose rates are calculated according to ACP source, shielding materials, etc., and reinforcement structures are determined considering the current situation of hot cells, installation of shield windows and the easiness of work.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.3
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• pp.1035-1039
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• 2010

Activation products generated by photoneutrons in concrete shielding wall around electron linac were estimated for a high energy X-ray container cargo inspection system. Monte carlo code, MCNPX2.5.0 was used for reference system of 9MeV fixed type dual-direction container cargo inspection system installed at major harbors in Korea. Activation products inventory generated by photoneutron (n,$\gamma$) reaction are estimated, and then radiation dose rate is calculated from the results.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.381-382
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• 2005

원자력발전소에서 발생하는 사용후연료 집합체를 운반하기 위한 수송용기는 고준위 방사성물질의 위험으로부터 인간과 환경을 보호하기 위하여 안전성이 철저하게 보장되어야만 한다. 원자력법과 IAEA 안전수송규정 등 국내외의 관련규정에 의하면 사용후연료 수송용기는 정상운반조건은 물론 수송 도중 발생할 수 있는 운반사고조건에서 B(U)F형 운반용기에 대한 기술기준을 만족시키어 어떠한 경우에도 방사선차폐, 임계, 격납, 열 및 구조적 건전성을 유지하여 방사성물질을 누출시키지 않아야 한다고 규정하고 있다. 본 논문은 한수원(주)에서 개발하여 현재 사용하고 있는 경수로형 사용후연료 수송용기(KN-12 수송용기)의 격납계통에 대한 건전성을 확인하기 위하여 해석에 의한 격납평가 및 수송용기의 운영 중 수행하는 누설시험 등의 누설평가방법에 대하여 기술하였으며, 또한, 매 운반 시 측정한 실제 누설률을 제시하고 분석하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analyses were performed for the container in a view of Type B. As a result of structural analysis, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at $134.8^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used.

• Journal of Radiation Protection and Research
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• v.11 no.1
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• pp.29-36
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• 1986

GESS-a computer code for simulation of energy spectra for gamma-ray in NaI(T1) scintillator has been developed. The Monte Carlo method was employed to simulate physical behaviours of particle transport in a medium. In the processes of simulation, all the interaction processes such as Rayleigh and Compton scattering, photoelectric effect and pair production were considered. The resulting electron slowing down spectrum was also considered with the CSDA model. For the purpose of verification of the code, a measurement gamma spectrum for incident gamma energy of 1.33 MeV was performed. The measured values appeared to be slightly higher than the theoretically calculated values.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.158-167
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• 2007

It is required that the neutron dosimetry be present to monitor the reactor vessel throughout its plant life. The Ex-vessel Neutron Dosimetry Systems which consist of sensor sets, radiometric monitors, gradient chains, and support hardware have been installed for 3-Loop plants after a complete withdrawal of all six in-vessel surveillance capsules. The systems have been installed in the reactor cavity annulus in order to characterize the neutron energy spectrum over the beltline region of the reactor vessel. The installed dosimetry were withdrawn and evaluated after a irradiation during one cycle and then compared to the cycle specific neutron transport calculations. The reaction rates from the measurement and calculation were compared and the results show good agreements each other.

• Journal of the Korean Society of Radiology
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• v.14 no.3
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• pp.193-201
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• 2020

The purpose of this study is to evaluate the photon characteristics according to the material and thickness of the electrons incidented through a linear accelerator. The computer simulation design is a linear accelerator target consisting of a 2 mm thick tungsten single material and a 1.8 mm and 2.3 mm thick tungsten and copper composite material. In the research method, First, the behavior of primary particles in the target was evaluated by electron fluence and electron energy deposition. Second, photons occurring within the target were evaluated by photon fluence. Finally, the photon angle-energy distribution at a distance of 1 m from the target was evaluated by photon fluence. As a result, first, electrons, which are primary particles, were not released out of the target for electron fluence and energy deposition in the target of a single material and a composite material. Then, electrons were linearly attenuated negatively according to the target thickness. Second, it was found that the composite material target had a higher photon generation than the single material target. This confirmed that the material composition and thickness influences photon production. Finally, photon fluence according to the angular distribution required for shielding analysis was calculated. These results confirmed that the photon generation rate differed depending on the material and thickness of the linear accelerator target. Therefore, this study is necessary for designing and operating a linear accelerator use facility for container security screening that is being introduced in the country. In addition, it is thought that it can be used as basic data for radiation protection.