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Relevant bibliographies by topics / Shielding (Radiation) – Evaluation

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Academic literature on the topic 'Shielding (Radiation) – Evaluation'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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Journal articles on the topic "Shielding (Radiation) – Evaluation"

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Cho, Yong-In, Chang-Soo Kim, and Jung-Hoon Kim. "Evaluation of syringe shield effectiveness in handling radiopharmaceuticals." Nuclear Technology and Radiation Protection 30, no. 2 (2015): 158–63. http://dx.doi.org/10.2298/ntrp1502158c.

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The purpose of this study was to evaluate the effectiveness of the radiation shield of radionuclide syringes and the personal dose equivalent by performing a simulation of radionuclides used in nuclear medicine diagnosis. In order to evaluate the dose depending on the distance between the radiation source and the ICRU sphere against the thickness of the shielding device, the distance at which a nuclear medicine worker may inadvertently come into contact with radiation from the radiation source was set at 0 cm to 30 cm according to the thickness of the shield, thus fixing the ICRU sphere. For a dose evaluation, Hp(10), Hp(3), and Hp(0.07) measurable in specific depth of the ICRU were evaluated. It was found that a dose measured on skin surface of nuclear medicine workers was relatively higher, that the dose varied in relation to the thickness of the radiation shield, and that the shielding effect decreased for some radiation sources such as 67Ga and 111In. It proved necessary to increase thickness of shielding device to the radiation sources such as 67Ga and 111In. It is also considered that a study of proper shielding thickness will be needed in future.

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CHIDA, KOICHI, YOSHIAKI MORISHIMA, YOSHIAKI KATAHIRA, HIROO CHIBA, and MASAYUKI ZUGUCHI. "Evaluation of Additional Lead Shielding in Protecting the Physician from Radiation during Cardiac Interventional Procedures." Japanese Journal of Radiological Technology 61, no. 12 (2005): 1632–37. http://dx.doi.org/10.6009/jjrt.kj00004022974.

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Sayyed, Mohammed I., Ferdi Akman, Veysel Turan, and Aslı Araz. "Evaluation of radiation absorption capacity of some soil samples." Radiochimica Acta 107, no. 1 (December 19, 2018): 83–93. http://dx.doi.org/10.1515/ract-2018-2996.

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Abstract The aim of the present work is to investigate the radiation absorption capacity of different soil samples in Turkey. For this purpose, we used a γ ray transmission geometry to measure the mass attenuation coefficients of eight soil samples collected between Bingöl city and Solhan district, Turkey at different γ-ray energies in the range of 13.94–88.04 keV. The radioactive sources utilized in the experiment were 241Am, 109Cd and 133Ba. FFAST and WinXCOM programs were used to evaluate the theoretical mass attenuation coefficients values of the selected soil samples. There is a good agreement between experimental and theoretical results. Additionally, the mass attenuation coefficients values used to evaluate different radiation shielding parameters such as effective atomic number, half value layer and mean free path. The variation of shielding parameters was examined for soil composition and photon energy. The obtained results revealed that S6 soil sample is the best soil in terms of shielding effectiveness among all the collected soils due to lower values for half value layer and mean free path. The effective removal cross-section (ΣR) of fast neutrons for the collected soils was also computed to examine neutrons shielding properties of the soil samples. It is found that the ΣR values for the soil samples are almost constant and lie within the range (0.04286–0.04949 cm−1).

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Krayushkin, V. V., P. A. Orlenko, and A. V. Larichev. "Technicoeconomic evaluation of radiation shielding for commercial electron accelerators." Soviet Atomic Energy 61, no. 3 (September 1986): 758–61. http://dx.doi.org/10.1007/bf01129883.

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Xu, Yun-Chuan, Quan-Ping Zhang, Jun-Hua Liu, You Wu, Li-Ping Liu, Dui-Gong Xu, and Yuan-Lin Zhou. "PbWO4 nanofibers for shielding gamma radiation: crystal growth, morphology and performance evaluation." CrystEngComm 20, no. 40 (2018): 6197–206. http://dx.doi.org/10.1039/c8ce01224e.

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Endo, A., and T. Sato. "Radiation transport calculations for cosmic radiation." Annals of the ICRP 41, no. 3-4 (October 2012): 142–53. http://dx.doi.org/10.1016/j.icrp.2012.06.010.

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The radiation environment inside and near spacecraft consists of various components of primary radiation in space and secondary radiation produced by the interaction of the primary radiation with the walls and equipment of the spacecraft. Radiation fields inside astronauts are different from those outside them, because of the body's self-shielding as well as the nuclear fragmentation reactions occurring in the human body. Several computer codes have been developed to simulate the physical processes of the coupled transport of protons, high-charge and high-energy nuclei, and the secondary radiation produced in atomic and nuclear collision processes in matter. These computer codes have been used in various space radiation protection applications: shielding design for spacecraft and planetary habitats, simulation of instrument and detector responses, analysis of absorbed doses and quality factors in organs and tissues, and study of biological effects. This paper focuses on the methods and computer codes used for radiation transport calculations on cosmic radiation, and their application to the analysis of radiation fields inside spacecraft, evaluation of organ doses in the human body, and calculation of dose conversion coefficients using the reference phantoms defined in ICRP Publication 110.

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Hossain, MN, MS Mia, TA Biman, H. Mehdi, and F. Begum. "Evaluation of the shielding parameters implemented in the PET-CT facility at the National Institute of Nuclear Medicine & Allied Sciences (NINMAS), Dhaka." Bangladesh Journal of Physics 26, no. 2 (September 20, 2020): 51–59. http://dx.doi.org/10.3329/bjphy.v26i2.49306.

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Positron Emission Tomography (PET) combined with Computed Tomography (CT) is a very sensitive diagnostic imaging modality that consists of both functional and anatomical imaging into one combined scanning system. Because of the high energy of annihilation radiation (511 keV) for PET tracers, shielding requirements are an important consideration in the design of a PET/CT facility. The shielding evaluation for adequate radiation protection of a PET/CT facility consists of the assessment of annual effective dose both to occupational workers and to members of the public’s. The overall shielding assessment takes into account the radionuclides activity involved, the facility design, the scanning procedures, the expected number of patients per year and so on. The evaluation also depends upon the size of the PET/CT imaging room and patient’s uptake rooms, the thickness and the physical materials of walls, floors and ceilings. In this work we verified the adequacy of shielding installed in the PET/CT facility at the National Institute of Nuclear Medicine & Allied Sciences (NINMAS), Dhaka. The assessment results were compared to the provided shielding which was the design requirements. Bangladesh Journal of Physics, 26(2), 51-59, December 2019

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YOSHIKAWA, Ema, Hideo KOMINE, Shigeru GOTO, Mitsugu YOSHIMURA, Akihiko SUZUKI, Seiichi NARUSHIMA, Yasunori ARAI, Shinsuke UJIIE, Yuki SAKODA, and Yasushi NAGAE. "THE QUANTITATIVE EVALUATION FOR RADIATION SHIELDING CAPABILITIES OF SOIL MATERIALS." Journal of Japan Society of Civil Engineers, Ser. C (Geosphere Engineering) 73, no. 4 (2017): 342–54. http://dx.doi.org/10.2208/jscejge.73.342.

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Saudi, H. A., and Ei-Said H. Ei-Mosallamy. "Evaluation of Polymethyl Methacrylate Sheets as a Radiation Shielding Material." Journal of Packaging Technology and Research 4, no. 3 (August 27, 2020): 227–33. http://dx.doi.org/10.1007/s41783-020-00098-6.

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Pomaro, B., V. A. Salomoni, F. Gramegna, G. Prete, and C. E. Majorana. "Radiation damage evaluation on concrete shielding for nuclear physics experiments." Annals of Solid and Structural Mechanics 2, no. 2-4 (November 3, 2011): 123–42. http://dx.doi.org/10.1007/s12356-011-0023-7.

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Dissertations / Theses on the topic "Shielding (Radiation) – Evaluation"

1

Bogue, Jonathan Nelson. "Evaluation of Patient-Scatter Factors for Radiation Therapy ShieldingUsing Physical Measurement in a "Good" Geometry." University of Toledo Health Science Campus / OhioLINK, 2018. http://rave.ohiolink.edu/etdc/view?acc_num=mco1525447575821026.

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Evangelides, George. "The analysis and evaluation of nuclear decay schemes for use in radiation shielding and other applications." Thesis, Imperial College London, 1988. http://hdl.handle.net/10044/1/47052.

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Savinov, Roman. "Evaluation of Efficiency of Various Materials to Shield From Radiation in Space Using the Monte Carlo Transport Code Called FLUKA." DigitalCommons@CalPoly, 2016. https://digitalcommons.calpoly.edu/theses/1687.

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The purpose of this study is to improve spacecraft shielding from radiation in space. It focuses on the evaluation of shielding efficiency of different materials. The efficiency of a shield is evaluated by the dose profile within the shield and the amount of dose absorbed by a target using the Monte Carlo transport code called FLUKA. The output of this code is validated by recreating the experiments from published papers and comparing the results. Once the FLUKA’s output is validated, the efficiency of sixteen materials, subject to SPE and GCR sources, are evaluated. The efficiency comparison is made by fixing the area density of a shield. It was found that polyethylene, water, carbon and silicon outperform aluminum – the primary metal used in spacecraft. In case of composite shield, made of layers of different materials, the 3Carb-9Al combination has better performance than the shield made just of aluminum. This holds true for both Solar Particle Events (SPEs) and Galactic Cosmic Ray (GCR). However, the choice of material is more efficient at shielding from SPE particles rather than from GCR. In case of GCR, the choice of materials is found to have rather small effect on the efficiency of a shield. The percent difference between the rate of dose absorption by a target, shielded by different materials, is within about 9%. Secondary particles make a significant contribution to the target’s dose. For SPEs, the secondary particles are primarily electrons and neutrons. For GCRs, the secondary particles are primarily pions, α-particles and electrons. Protons contribute more than 50% to the target’s dose in both cases.

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Kildea, John. "An evaluation of NCRP Report No. 151 - radiation shielding design for radiotherapy facilities, and a feasibility study for 6 MV open-door treatments in an existing high-energy radiation therapy bunker." Thesis, McGill University, 2010. http://digitool.Library.McGill.CA:8881/R/?func=dbin-jump-full&object_id=92271.

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MOMESSO, ROBERTA G. R. A. P. "Desenvolvimento e validação de um referencial metodológico para avaliação da cultura de segurança de organizações nucleares." reponame:Repositório Institucional do IPEN, 2017. http://repositorio.ipen.br:8080/xmlui/handle/123456789/28035.

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Submitted by Pedro Silva Filho (pfsilva@ipen.br) on 2017-11-22T16:34:17Z No. of bitstreams: 0
Made available in DSpace on 2017-11-22T16:34:17Z (GMT). No. of bitstreams: 0
A cultura de segurança na área nuclear é definida como o conjunto de características e atitudes da organização e dos indivíduos que fazem que, com uma prioridade insuperável, as questões relacionadas à proteção e segurança nuclear recebam a atenção assegurada pelo seu significado. Até o momento, não existem instrumentos validados que permitam avaliar a cultura de segurança na área nuclear. Em vista disso, os resultados da definição de estratégias para o seu fortalecimento e o acompanhamento do desempenho das ações de melhorias tornam-se difíceis de serem avaliados. Este trabalho teve como objetivo principal desenvolver e validar um instrumento para a avaliação da cultura de segurança de organizações nucleares, utilizando o Instituto de Pesquisas Energéticas e Nucleares como unidade de pesquisa e coleta de dados. Os indicadores e variáveis latentes do instrumento foram definidos utilizando como referência modelos de avaliação de cultura de segurança da área da saúde e área nuclear. O instrumento de coleta de dados proposto inicialmente foi submetido à avaliação por especialistas da área nuclear e, posteriormente, ao pré-teste com indivíduos que pertenciam à população pesquisada. A validação do modelo foi feita por meio da modelagem por equações estruturais utilizando o método de mínimos quadrados parciais (Partial Least Square - Structural Equation Modeling PLS-SEM), no software SmartPLS. A versão final do instrumento foi composta por quarenta indicadores distribuídos em nove variáveis latentes. O modelo de mensuração apresentou validade convergente, validade discriminante e confiabilidade e, o modelo estrutural apresentou significância estatística, demonstrando que o instrumento cumpriu adequadamente todas as etapas de validação.
Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

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Kegopotsemang, Onalenna. "Evaluation and redesign of radiation shielding in a radionuclide production facility at a particle accelerator / Onalenna Kegopotsemang." Thesis, 2004. http://hdl.handle.net/10394/11338.

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iThemba LABS is a particle accelerator facility housing a radionuclide production facility that uses a 66 MeV proton beam to produce radionuclides for medical and industrial use. Ionising radiation is produced by a variety of sources at Themba LABS. Ionising is a health hazard. High doses can cause acute radiation syndrome, i.e. "radiation sickness". Lower doses cannot cause acute symptom, but carry a risk of radiation-related cancer. Ionising radiation is also detrimental to materials, and can damage polymers and lubricants e.g. Shielding is used to reduce radiation levels to values that should be safe for the intended level of human occupancy. Shield performance is vital to human health and the life expectancy of polymers and lubricants, so that quality management in shield design is very important. However, until 2003, there has been no formal system at iThemba LABS to evaluate and improve all radiation shield designs and layouts from a radiation protection perspective. This study deals with evaluating and redesigning radiation shielding in the radionuclide production facility of iThemba LABS. There are several designs and layouts in the radionuclide production building of iThemba LABS that lead to unnecessary exposure of personnel to ionising radiation. The shielding in these areas are sub-standard. Performance criteria for radiation shields are developed. Inadequate radiation shields are identified. The identified inadequate shields are: the processing hotcells, the target store room and the hotcells in the radiopharmaceutical dispensing laboratory, Point-Kernel radiation shielding calculations are done to specify materials and material thickness that will adequately protect workers against ionising radiation.
Thesis (MSc. ARST) North-West University, Mafikeng Campus, 2004

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Thompson, Kyle Richard. "Modeling of gamma rays streaming through straight rectangular ducts." 1985. http://hdl.handle.net/2097/27559.

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Worn, Cheyn. "An Evaluation of Shadow Shielding for Lunar System Waste Heat Rejection." Thesis, 2012. http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-11202.

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Shadow shielding is a novel and practical concept for waste heat rejection from lunar surface spacecraft systems. A shadow shield is a light shield that shades the radiator from parasitic thermal radiation emanating from the sun or lunar surface. Radiator size and mass can reduce if the radiator is not required to account for parasitic heat loads in addition to system energy rejection requirements. The lunar thermal environment can be very harsh towards radiative heat rejection. Parasitic heat loads force the radiator to expand in size and mass to compensate. On the Moon, there are three types: surface infrared, solar insulation, and albedo. This thesis tests shadow shielding geometry and its effect on the radiator and nuclear reactor in a reactor-powered Carnot heat engine. Due to the nature of cooling by radiative heat transfer, the maximum shaft work a Carnot system can produce and the minimal required radiator area occurs when the Carnot efficiency is 25%. First, a case for shadow shielding is made using an isothermal, control radiator model in Thermal Desktop. Six radiator temperatures and three latitudes are considered in the tests. Test variables in this section include radiator shapes and shade geometry. The simulations found that shadow shielding is best suited for a low-temperature radiator at the lunar equator. Optimized parabolic shade geometry includes a focus right above or at the top of the radiator and full to three-quarters shade height. The most useful rectangular radiator shape for shadow shielding is that which has a low height and long width. All simulations were conducted using a shade with a 10 kg/m2 area mass. A sensitivity study was conducted for different shade area masses using high and low values found in the literature. The shade is the most useful when the shade's area mass is less than or equal to that of the radiator. If the shade mass is below this threshold, the shade would be applicable to all radiator temperatures tested. Optimized shade and radiator geometry results were then factored into a second model where the radiator is comprised of heat pipes which is similar to radiators from actual system designs. Further simulations were conducted implementing the SAFE-4001 fast fission nuclear reactor design. The study found that shadow shielding allowed the system to use a low-temperature radiator where other configurations were not viable because shadow shielding drastically improves radiative heat transfer from the radiator, but at the consequence of raising radiator mass.

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Books on the topic "Shielding (Radiation) – Evaluation"

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W, Wilson John. NUCRFG2: an evaluation of the semiempirical nuclear fragmentation database. Hampton, Va: Langley Research Center, 1995.

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Structural Shielding Design And Evaluation for Megavoltage X-and Gamma-ray Radiotherapy Facilities: Recommendations of the National Council on Radiation ... And Measurements (Ncrp Report Series). National Council of Teachers of English, 2005.

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Conference papers on the topic "Shielding (Radiation) – Evaluation"

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Acedillo, Shannen Moira M. "Evaluation of the Application of Carbon Nanotubes for Radiation Shielding." In AIAA SPACE 2015 Conference and Exposition. Reston, Virginia: American Institute of Aeronautics and Astronautics, 2015. http://dx.doi.org/10.2514/6.2015-4505.

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Nounu, Hatem, Myung-Hee Kim, and Francis Cucinotta. "Radiation Shielding Evaluation Tools for Risk Reduction on Future Human Space Missions." In 42nd International Conference on Environmental Systems. Reston, Virigina: American Institute of Aeronautics and Astronautics, 2012. http://dx.doi.org/10.2514/6.2012-3645.

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Kaur, Parminder, K. J. Singh, and Sonika Thakur. "Evaluation of the gamma radiation shielding parameters of bismuth modified quaternary glass system." In 2ND INTERNATIONAL CONFERENCE ON CONDENSED MATTER AND APPLIED PHYSICS (ICC 2017). Author(s), 2018. http://dx.doi.org/10.1063/1.5032878.

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Takada, Shoji, Shunki Yanagi, Kazuhiko Iigaki, Masanori Shinohara, Daisuke Tochio, Yosuke Shimazaki, Masato Ono, and Kazuhiro Sawa. "Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16095.

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HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

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Hashimoto, Tomoharu, Masahiro Kondo, Ryuichi Tayama, and Hideho Gamo. "Development of New Simulation Software System to Evaluate Radiation Dose Rates in a Wide Radioactively Contaminated Area." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-55069.

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The Japanese government plans to conduct decontamination tasks in radioactively contaminated areas. For such a situation, we developed a system that evaluates radiation dose rates in a wide radioactively contaminated area by utilizing our radiation dose evaluation technology. This system can not only generate present maps of radiation dose rate in the air based on the dose rate measured at the surface of the contaminated areas, but can also quickly calculate the reduction effect of dose rate due to decontamination tasks by entering decontamination factors. The system can then formulate decontamination plans and make it possible to plan measures to reduce radiation exposure for workers and local residents. Radioactive nuclides that contribute to gamma-ray dose rate are mainly Cs-134 and Cs-137 in soil, on trees, buildings, and elsewhere. Shapes of such radiation sources are assumed to be 10m square or 100m square. If it is unsuitable that the radiation sources assume to squares, the radiation sources can assume to point. The relation between distance from the surface or point source and the radiation dose rate is calculated using MCNP5 code (A General Monte Carlo N-Particle Transport Code - Version 5), and approximated using four-parameter empirical formula proposed by Harima et al. In addition, the system can consider shielding such as soil, concrete, and iron. When setting such shielding, the skyshine dose rate is taken into account in dose rate calculation.

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Sanfiorenzo, Antonio, Rosa Lo Frano, Giovanni Pugliese, and Giuseppe Forasassi. "Horizontal Drop Test Evaluation of a Packaging System." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-31201.

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In the transportation of radioactive waste, the package is designed as the major engineered system capable to ensure the containment and provide safety functions, such as radiation shielding, structural integrity against external mechanical and thermal loads, dissipation of the decay heat, etc. Packaging systems are designed in accordance to rigorous acceptance requirements, like the International Atomic Energy Agency (IAEA) ones, so to provide protection to human being and environment against radiation exposure and contamination, particularly in reference accident scenarios including, as it is widely known in literature, drop, puncture, fire and submersion tests. The scope of the present study is to evaluate the structural response and performance in a free drop test condition of a new Italian packaging system that should be used for the transportation of low and intermediate level radioactive wastes. For this purpose the carried out numerical analyses are presented and discussed. The numerical analyses, performed by the finite element MARC® code, simulate the behaviour of the packaging system components: the overpack, gasket, cover lid, bolts and a concrete matrix representative of the radioactive content. The obtained results for 1.2 m horizontal drop, on a flat and unyielding surface, were critically analysed and also compared to the experimental ones obtained from the experimental test campaign performed at the Unipi test facility on the new Italian packaging system considered. The stress and acceleration values indicate that the package, although rather local deformations in correspondence of bolts and secondary lid, is capable to withstand the dynamic loading generated during the drop test without any unacceptable loss of the safety features.

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Sun, Sida, Sheng Fang, and Hong Li. "Radiation Shielding Design of High Temperature Reactor Pebble-Bed Module (HTR-PM)." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60413.

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Radiation safety is an important concern in the design and licensing of the 200MWe High Temperature Reactor Pebble-bed Module (HTR-PM) demonstration power plant in China. To meet the requirement of the regulatory, various radiation protection strategies and methods are applied in the design process of systems and components of HTR-PM. In this study, the radiation shielding design of HTR-PM is reviewed, which includes the radiation source analysis, in-house dose calculation tool, shielding and dose reduction methods used for primary systems. The underlying conservative assumption is also discussed for correctly evaluating the dose calculation result. This summary provides a relatively systematic review of the radiation shielding methods in the design phase of HTR-PM, which may provide useful information and experiences for the radiation shielding design of future pebble-bed reactors.

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Nogueron, Jose A., Zhang Baorui, and Zhou Zhiwei. "Optimal Design for Helium Cooled Solid Breeder Blanket of CFETR." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16345.

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Abstract On the road to achieving fusion energy production at a commercial level, China has proposed the design of a nuclear fusion reactor based on the tokamak configuration. Aimed to produce self-sustained burning plasma and a closed tritium breeding cycle, this device is expected to provide a bridge between ITER and DEMO designs. The Breeding Blanket (BB) is one of the key technological challenges to be designed in order to guarantee sufficient tritium production, heat removal capabilities and radiation shielding protection. Two preliminary designs of the Helium-Cooled Solid Breeder (HCSB) blanket have been suggested to be subjected to further investigation. Despite the fact that both designs use the same materials, they present completely different geometrical arrangements. In the present analysis, these two concepts of the HCSB are examined, addressing the critical design issues that affect the performance of the system. A neutronic analysis is performed to calculate the Tritium Breeding Ratio (TBR), which is a crucial parameter to meet tritium production requirements. Based on this investigation, the cooling capacity of the reactor is analyzed with the help of a Computational Fluid Dynamic (CFD) software through a dynamic evaluation. Valuable conclusions can be extracted from the results of this work, which can be referenced for investigations regarding further studies of HCSB blankets.

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Kern, John W. "Evaluating Spacecraft Radiation Shielding Using a Microcomputer-Based CAD Program." In International Conference On Environmental Systems. 400 Commonwealth Drive, Warrendale, PA, United States: SAE International, 1992. http://dx.doi.org/10.4271/921371.

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Kimura, Ken-Ichi, Akira Hasegawa, Katsumi Hayashi, Mikio Uematsu, Tomohiro Ogata, Takao Tanosaki, Ryoetsu Yoshino, Mituru Sato, Minoru Saito, and Masaharu Kinno. "Development of Low-Activation Design Method for Reduction of Radioactive Waste Below Clearance Level." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48484.

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Design methodology for reinforced concrete of nuclear power plants to reduce radioactive wastes in decommission phase has been developed. To realize this purpose, (1) development of raw materials database of cements, aggregates and steel bars on concentration of radioactive target elements, (2) trial production of low activation cements and steel bars based on the material database developed in (1), and (3) development of tools for estimation and prediction of the amount of radioactive elements in reactor shielding walls have been carried out. Radioactive analysis showed that Co and Eu were the major target elements which decide the radioactivity level of reinforced concrete from wide survey of raw materials for concrete (typically aggregates and cements). Material database for the contents of Co and Eu was developed based on the chemical analysis and radioactivation analysis. Upon the above survey and execution expreiment of concrete, six types of low-activation concrete are proposed for various radioactive portion in the plant. These concrete have a 1/10 – 1/300 rasioactivity compare to the ordinary concrete, which are assumed the concrete with Andesite aggregate and ordinary Portland cement. Baed on the above data base, it was clarified that the low activation cement would be successfully manufactured by adequate selection of raw materials. The prospect to produce the low-heat portland cement which would have a 1/3 radioactivity in comparison with conventioanl cements obtained by means of selection of limestone and natural gypsum. An attempte was carried out to produece low activation heavy-mortar which would have radioactivity below the clearance level when using at the radiation shielding wall of BWR. Characterization and optimization of consturction conditions with new additives have also been carried out. These two new raw materials for low-activation concrete are conducted in pre-manufacture size, and over the laboratry level. Boron added low-activation concrete are also carried out as extreamly high performance low-activation concrete. It was claryfied that the accurcy of calculation results of the radioactivity evaluation was very high compared to available benchmark calculation for the JPDR and commercial light water reactor. The specification of the mapping system for judging the activation classification was also developed by using the general-purpose radio activation calculation tool. This work is supported by a grant-in-aid of Innovative and Viable Nuclear Technology (IVNET) development project of Ministry of Economy, Trade and Industry, Japan.

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