• Title/Summary/Keyword: Shielding Evaluation

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RADIATION SHIELDING EVALUATION OF IP-2 PACKAGES FOR LOW- AND INTERMEDIATE-LEVEL RADIOACTIVE WASTE

  • Kim, Min-Chul;Choi, Jong-Rak;Chung, Sung-Hwan;Ko, Jae-Hoon
    • Nuclear Engineering and Technology
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    • v.40 no.6
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    • pp.511-516
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    • 2008
  • Korea Hydro & Nuclear Power Co., Ltd. (KHNP) developed new IP-2 packages to transport low- and intermediate-level radioactive waste (LILW) steel drums from nuclear power plants to a disposal facility in accordance with IAEA and Korean transport regulations of radioactive material. Radiation shielding evaluation of the packages was carried out to demonstrate compliance with the regulatory requirements for IP-2 packages of radioactive material. Dose rate limits of LILW drums contained in the packages were determined.

The Evaluation of Minimum Cooling Period for Loading of PWR Spent Nuclear Fuel of a Dual Purpose Metal Cask (국내 경수로 사용후핵연료의 금속 겸용용기 장전을 위한 최소 냉각기간 평가)

  • Dho, Ho-Seog;Kim, Tae-Man;Cho, Chun-Hyung
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.14 no.4
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    • pp.411-422
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    • 2016
  • Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

Monte Carlo shielding evaluation of a CSNS Multi-Physics instrument

  • Liang, Tairan;Shen, Fei;Yin, Wen;Xu, Juping;Yu, Quanzhi;Liang, Tianjiao
    • Nuclear Engineering and Technology
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    • v.51 no.8
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    • pp.1998-2004
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    • 2019
  • The Multi-Physics (MP) instrument is one of 20 neutron spectrometers planned in the China Spallation Neutron Source (CSNS). This paper presents a shielding calculation for the MP instrument using Monte Carlo codes MCNPX and FLUKA. First, the neutrons that escape from the CSNS decoupled water moderator and are delivered to the beam line of the MP instrument are calculated to use as the source term of the shielding calculation. Then, to validate the calculation method based on multiple variance reduction techniques, a cross check between MCNPX and FLUKA codes is performed by comparing the calculation results of the dose rate distribution on a simplified beam line model. Finally, a complete geometry model of the MP instrument is set up, and the primary parameters for the shielding design are obtained according to the calculated dose rate map considering different worst-case scenarios.

Literature Review on Material Development and Performance Evaluation Method for EMP Shielding Concrete (EMP 차폐 콘크리트 개발 및 성능평가 방법에 관한 문헌 연구)

  • Lee, Woong-Jong;Lee, Hwan;Kim, Young-Jin
    • Journal of the Korea Academia-Industrial cooperation Society
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    • v.21 no.12
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    • pp.67-76
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    • 2020
  • The purpose of this study was to derive the directionality of technology development of high-power electromagnetic pulse (EMP) shielding concrete and standardization of a shielding performance evaluation method. Because the EMP shielding mechanism of concrete has not been identified clearly, and the verification method for EMP shielding performance has not been standardized, it is difficult to compare the research results between researchers. The development direction of EMP shielding concrete was derived from a consideration of the electromagnetic wave loss mechanism of metal. The standardization direction for verifying the EMP shielding performance of concrete was derived from a consideration of the electrical properties of concrete and the shielding performance evaluation methods of previous studies. As a result, the development of electrically conductive concrete is required, and test methods classified by the electromagnetic wave loss mechanism should be applied. For quality verification, the development of EMP shielding concrete will be feasible and its performance can be evaluated if a test method referencing the generalized shielding evaluation method (MIL-STD, etc.) is applied.

SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS

  • Ko, Jae-Hun;Park, Jea-Ho;Jung, In-Soo;Lee, Gang-Uk;Baeg, Chang-Yeal;Kim, Tae-Man
    • Nuclear Engineering and Technology
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    • v.46 no.4
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    • pp.547-556
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    • 2014
  • Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

Development for Improvement Methodology of Radiation Shielding Evaluation Efficiency about PWR SNF Interim Storage Facility (PWR 사용후핵연료 중간저장시설의 몬테칼로 차폐해석 방법에 대한 계산효율성 개선방안 연구)

  • Kim, Taeman;Seo, Myungwhan;Cho, Chunhyung;Cha, Gilyong;Kim, Soonyoung
    • Journal of Radiation Protection and Research
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    • v.40 no.2
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    • pp.92-100
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    • 2015
  • For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.