• Title/Summary/Keyword: Neutron Dose Assessment

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An Assessment of the Secondary Neutron Dose in the Passive Scattering Proton Beam Facility of the National Cancer Center

  • Han, Sang-Eun;Cho, Gyuseong;Lee, Se Byeong
    • Nuclear Engineering and Technology
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    • v.49 no.4
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    • pp.801-809
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    • 2017
  • The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a $^3He$ neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from $4.942{\pm}0.031mSv/Gy$ at the end of the field to $0.324{\pm}0.006mSv/Gy$ at 150 cm in axial distance.

APPLICATION OF WHOLE BODY COUNTER TO NEUTRON DOSE ASSESSMENT IN CRITICALITY ACCIDENTS

  • Kurihara, O.;Tsujimura, N.;Takasaki, K.;Momose, T.;Maruo, Y.
    • Journal of Radiation Protection and Research
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    • v.26 no.3
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    • pp.249-253
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    • 2001
  • Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of $^{24}Na$ in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of $^{24}Na$ is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 $[(Bq^{24}Na/g^{23}Na)/mGy]$.

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Occupational radiation exposure control analyses of 14 MeV neutron generator facility: A neutronic assessment for the biological and local shield design

  • Swami, H.L.;Vala, S.;Abhangi, M.;Kumar, Ratnesh;Danani, C.;Kumar, R.;Srinivasan, R.
    • Nuclear Engineering and Technology
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    • v.52 no.8
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    • pp.1784-1791
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    • 2020
  • The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 1012 n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

AN ASSESSMENT OF THE RADIATION DOSE RATE DUE TO AN OCCURRENCE OF THE DEFECT ON THE SPENT NUCLEAR FUEL ROD

  • Lee, Sang-Hun;Moon, Joo-Hyun
    • Journal of Radiation Protection and Research
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    • v.34 no.3
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    • pp.144-150
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    • 2009
  • This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

Assessment of Neutron Skyshine Dose in a Cargo Inspection Facility Using High Energy X-ray (고에너지 X-ray를 이용한 화물검색시설에서의 중성자 Skyshine 방사선량률 평가)

  • Cho, Young-Ho
    • Journal of the Korean Society of Radiology
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    • v.2 no.3
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    • pp.27-31
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    • 2008
  • The radiation protection measures for the photoneutrons are one of the most important issue of radiation safety in high energy X-ray facilities. When the photoneutrons are released from the facility, the general public as well as occupational workers are exposed to unexpected radiations by neutron skyshine effect. In this study, the photoneutron inventory are calculated using monte carlo mothed, and the neutron skyshine dose rate is assessed using the inventory. A 9MeV X-ray cargo inspection facility is considered as a reference facility.

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Preliminary Study on Applicability of Accumulate Personal Neutron Dosimeter for Cosmic-ray Exposure of Aviators (운항승무원의 우주방사선 피폭 평가에 있어 누적형 개인 중성자 선량계의 적용가능성 예비 연구)

  • Kim, Hyeong-Jin;Chang, Byung-Uck;Byun, Jong-In;Song, Myeong Han;Kim, Jung-Ho
    • Journal of Radiation Protection and Research
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    • v.38 no.1
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    • pp.44-51
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    • 2013
  • ICRP recommended that cosmic ray exposure to the pilot and cabin crew would be considered as an occupational exposure due to their relatively high exposure. Since 2012 with the Act No. 10908 (Natural radiation management), the guideline of cosmic ray exposure to the pilot was established in Korea. The applicability of the solid-state nuclear track detector for personal dose assessment of pilot and cabin crew was evaluated. Dose linearity and angle dependence of dosimeters to the neutron were evaluated by $^{252}Cf$ neutron emitting source. The track density has a good agreement with the dose ($r^2$=0.99) and highly dependent on the degree of an angular of the dosimeter to the neutron source. In addition, the dosimeters (SSNTD) were exposed to cosmic ray in an aircraft during its cruising for more than two months in collaboration with Airline Pilots Association of Korea. Although the correlation between the track density from aircraft cruising altitude and expected neutron dose is low, however RSNS dosimeter could be used for personal neutron dosimeter. For application of RSNS as a personal dosimeter for pilot and cabin crew, additional studies are required.

Experimental Study on the Determination of Absorbed dose Index (흡수선량지수결정(吸收線量指數決定)에 관한 실험적(實驗的) 연구(硏究))

  • Jun, Jae-Shik;Rho, Chae-Shik;Ro, Seung-Gy;Ha, Chung-Woo;Yoo, Young-Soo;Lee, Hyun-Duk
    • Journal of Radiation Protection and Research
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    • v.7 no.1
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    • pp.34-48
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    • 1982
  • The prime purpose of this study is to realize an index quantity, absorbed dose index, defined by the ICRU for the characterization of ambient radiation level at any location for the purpose of radiation protection. The experiment has been designed to be carried out in two phases, namely, preliminary and main experiment. In the primary study a 30cm diameter sphere of polyethylene was used, while in the main experiment that of tissue equivalent material was fabricated and used. Both experiments were performed in the gamma-ray fields of $^{137}Cs\;and\;^{60}Co$, and in a neutron beam of thermal column of the TRIGA MARK-II research reactor. In the measurement of gamma-ray absorbed dose TLD-700 $(^{7}LiF)$ chips were used, and for the neutron dose both Au activation foils and TLD chips (TLD-600 $(^{6}LiF)$ and TLD-700 for the discrimination of gamma-ray contribution) were used. Theoretical assessment of the absorbed dose in the sphere phantom has been carried out in accordance with the Ehrlich's idea that deduced on the basis of Burlin's cavity theory in the case of gamma-ray irradiation. For the analysis of neutron dose fluence-KERMA rate conversion method was used. The explanation on the dose assessment is given in detail. Results obtained were numerically and statistically analyzed and the depth dose distributions are presented in the graphic forms with normalized values. In the concluding remarks, the possibility and difficulty of realizing the index quantity, including questions and problems to be solved are mentioned.

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A Study on the Evaluation of Surface Dose Rate of New Disposal Containers Though the Activation Evaluation of Bio-Shield Concrete Waste From Kori Unit 1

  • Kang, Gi-Woong;Kim, Rin-Ah;Do, Ho-Seok;Kim, Tae-Man;Cho, Chun-Hyung
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.19 no.1
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    • pp.133-140
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    • 2021
  • This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g-1 60Co and 2.63×105 Bq·g-1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr-1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr-1) as per global standards.

Shielding analyses supporting the Lithium loop design and safety assessments in IFMIF-DONES

  • Gediminas Stankunas ;Yuefeng Qiu ;Francesco Saverio Nitti ;Juan Carlos Marugan
    • Nuclear Engineering and Technology
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    • v.55 no.4
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    • pp.1210-1217
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    • 2023
  • The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

Demonstration of the Effectiveness of Monte Carlo-Based Data Sets with the Simplified Approach for Shielding Design of a Laboratory with the Therapeutic Level Proton Beam

  • Lai, Bo-Lun;Chang, Szu-Li;Sheu, Rong-Jiun
    • Journal of Radiation Protection and Research
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    • v.47 no.1
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    • pp.50-57
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    • 2022
  • Background: There are several proton therapy facilities in operation or planned in Taiwan, and these facilities are anticipated to not only treat cancer but also provide beam services to the industry or academia. The simplified approach based on the Monte Carlo-based data sets (source terms and attenuation lengths) with the point-source line-of-sight approximation is friendly in the design stage of the proton therapy facilities because it is intuitive and easy to use. The purpose of this study is to expand the Monte Carlo-based data sets to allow the simplified approach to cover the application of proton beams more widely. Materials and Methods: In this work, the MCNP6 Monte Carlo code was used in three simulations to achieve the purpose, including the neutron yield calculation, Monte Carlo-based data sets generation, and dose assessment in simple cases to demonstrate the effectiveness of the generated data sets. Results and Discussion: The consistent comparison of the simplified approach and Monte Carlo simulation results show the effectiveness and advantage of applying the data set to a quick shielding design and conservative dose assessment for proton therapy facilities. Conclusion: This study has expanded the existing Monte Carlo-based data set to allow the simplified approach method to be used for dose assessment or shielding design for beam services in proton therapy facilities. It should be noted that the default model of the MCNP6 is no longer the Bertini model but the CEM (cascade-exciton model), therefore, the results of the simplified approach will be more conservative when it was used to do the double confirmation of the final shielding design.