Non-combustible nuclear radiation shields with high hydrogen content

International Nuclear Information System (INIS)

Hall, W.C.; Peterson, J.M.

1978-01-01

The invention relates to compositions, methods of production, and uses of non-combustible nuclear radiation shields, with particular emphasis on those containing a high concentration of hydrogen atoms, especially effective for moderating neutron energy by elastic scatter, dispersed as a discontinuous phase in a continuous phase of a fire resistant matrix

Shielding behavior of multi-transformation phase change materials (MTPCM) against nuclear radiations

International Nuclear Information System (INIS)

Kumar, Ravindra; Goplani, Deepak; Kumar, Rohitash; Das, Mrinal Kumar; Kumar, Pramod; Jodha, Ajay Singh; Misra, Manoj; Khatri, P.K.

2008-01-01

In nuclear hardened structures and AFV's, special shielding materials are being used to provide protection from radiations generated in nuclear blast. However, in blast an intense heat pulse is also generated along with radiation. Currently used shield does not take care of this heat pulse. Defence Laboratory, Jodhpur has developed multi transformation phase change materials (MTPCM) based cool panels for passive moderation of temperature in severe desert heat. The MTPCM contains light nuclei of hydrogen, carbon and oxygen, and thus can absorb good amount of neutrons. MTPCM can also absorb intense heat pulse along with heat generated by secondary fires during blast as its latent heat (160-170 J/g) without significant rise in temperature (melting point 36-38 deg. C). Thus MTPCM can provide protection against both radiation as well as heat pulse generated in a nuclear blast along with its designed regular function of passively moderating temperature below 40 deg C during severe desert summer. A study has been undertaken to explore multiple applications of MTPCM panel. Protection factor provided by standard MTPCM panels against neutron and gamma radiations (both initial and fall out) were measured and results compared with PF provided by special lining pad currently being used in AFV's and field structures for nuclear protection. It is observed that MTPCM provides good PF (2.17) against neutron which is better than currently used shield pads (PFP%1.8). Present paper discusses results of this study. (author)

Radiation shielding performance of some concrete

International Nuclear Information System (INIS)

Akkurt, I.; Akyildirim, H.; Mavi, B.; Kilincarslan, S.; Basyigit, C.

2007-01-01

The energy consumption is increasing with the increased population of the world and thus new energy sources were discovered such as nuclear energy. Besides using nuclear energy, nuclear techniques are being used in a variety of fields such as medical hospital, industry, agriculture or military issue, the radiation protection becomes one of the important research fields. In radiation protection, the main rules are time, distance and shielding. The most effective radiation shields are materials which have a high density and high atomic number such as lead, tungsten which are expensive. Alternatively the concrete which produced using different aggregate can be used. The effectiveness of radiation shielding is frequently described in terms of the half value layer (HVL) or the tenth value layer (TVL). These are the thicknesses of an absorber that will reduce the radiation to half, and one tenth of its intensity respectively. In this study the radiation protection properties of different types of concrete will be discussed

Heavy density concrete for nuclear radiation shielding and power stations: [Part]2

International Nuclear Information System (INIS)

Singha Roy, P.K.

1987-01-01

This article is the second part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. In this part, some of the important properties of heavy density concrete are discussed. They include density, water retentivity, air content, permeability with special reference to concrete mixes used in India's nuclear power reactors. All these properties are affected to various extents by heating. Indian shield concrete is rarely subjected to temperatures above 60degC during its life, because of thermal shield protection. During placement, the maximum anticipated rise in temperature due to heat of hydration is restricted to around 45degC by chilling, if necessary to reduce shrinkage stresses and cracks. (M.G.B.)

Radiation shielding

International Nuclear Information System (INIS)

Yue, D.D.

1979-01-01

Details are given of a cylindrical electric penetration assembly for carrying instrumentation leads, used in monitoring the performance of a nuclear reactor, through the containment wall of the reactor. Effective yet economical shielding protection against both fast neutron and high-energy gamma radiation is provided. Adequate spacing within the assembly allows excessive heat to be efficiently dissipated and means of monitoring all potential radiation and gas leakage paths are provided. (UK)

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

E-Alerts: Nuclear science and technology (radiation shielding, protection, and safety). E-mail newsletter

International Nuclear Information System (INIS)

1999-01-01

Topics include: Shielding design, nuclear radiation transport properties of materials, decontamination; Container design and transportation requirements for radioactive materials; and Fallout shelters

Design experience: CRBRP radiation shielding

International Nuclear Information System (INIS)

Disney, R.K.; Chan, T.C.; Gallo, F.G.; Hedgecock, L.R.; McGinnis, C.A.; Wrights, G.N.

1978-11-01

The Clinch River Breeder Reactor Plant (CRBRP) is being designed as a fast breeder demonstration project in the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program. Radiation shielding design of the facility consists of a comprehensive design approach to assure compliance with design and government regulatory requirements. Studies conducted during the CRBRP design process involved the aspects of radiation shielding dealing with protection of components, systems, and personnel from radiation exposure. Achievement of feasible designs, while considering the mechanical, structural, nuclear, and thermal performance of the component or system, has required judicious trade-offs in radiation shielding performance. Specific design problems which have been addressed are in-vessel radial shielding to protect permanent core support structures, flux monitor system shielding to isolate flux monitoring systems for extraneous background sources, reactor vessel support shielding to allow personnel access to the closure head during full power operation, and primary heat transport system pipe chaseway shielding to limit intermediate heat transport system sodium system coolant activation. The shielding design solutions to these problems defined a need for prototypic or benchmark experiments to provide assurance of the predicted shielding performance of selected design solutions and the verification of design methodology. Design activities of CRBRP plant components an systems, which have the potential for radiation exposure of plant personnel during operation or maintenance, are controlled by a design review process related to radiation shielding. The program implements design objectives, design requirements, and cost/benefit guidelines to assure that radiation exposures will be ''as low as reasonably achievable''

Calculation of shielding and radiation doses for PET/CT nuclear medicine facility

International Nuclear Information System (INIS)

Mollah, A.S.; Muraduzzaman, S.M.

2011-01-01

Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)

Specialty glass development for radiation shielding windows and nuclear waste immobilization

International Nuclear Information System (INIS)

Mandal, S.; Ghorui, S.; Roy Chowdhury, A.; Sen, R.; Chakraborty, A.K.; Sen, S.; Maiti, H.S.

2015-01-01

The technology of two important varieties of specialty glasses, namely high density Radiation Shielding Window (RSW) glass and specialty glass beads of borosilicate composition have been successfully developed in CGCRI with an aim to meet the countries requirement. Radiation Shielding Windows used in nuclear installations, are viewing devices, which allow direct viewing into radioactive areas while still providing adequate protection to the operating personnel. The glass blocks are stabilized against damage from radiation by introducing cerium in definite proportions. Considering the essentially of developing an indigenous technology to make the country self-sufficient for this critical item, CGCRI has taken up a major programme to develop high lead containing glasses required for RSWs under a MoD with BARC. On the other hand, the specialty glass bead of specific composition and properties is a critical material required for management of radioactive waste in a closed nuclear fuel cycle that is followed by India. During reprocessing of the spent nuclear fuel, high level radio-active liquid waste (HLW) is produced containing unwanted radio isotopes some of which remain radioactive for thousands of years. The need is to immobilize them within a molecular structure so that they will not come out and be released to the ambience and thereby needs to be resolved if nuclear power is to make a significant contribution to the country's power requirement. Borosilicate glass has emerged as the material of choice for immobilization due to its unique random network structure

Radiation shielding application of lead glass

International Nuclear Information System (INIS)

Nathuram, R.

2017-01-01

Nuclear medicine and radiotherapy centers equipped with high intensity X-ray or teletherapy sources use lead glasses as viewing windows to protect personal from radiation exposure. Lead is the main component of glass which is responsible for shielding against photons. It is therefore essential to check the shielding efficiency before they are put in use. This can be done by studying photon transmission through the lead glasses. The study of photon transmission in shielding materials has been an important subject in medical physics and is potential useful in the development of radiation shielding materials

Status of the ITER tokamak nuclear shielding and radiological protection design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

2016-11-01

Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

Recent trends in radiation shielding: a RSIC perspective

International Nuclear Information System (INIS)

Trubey, D.K.; Roussin, R.W.; Maskewitz, B.F.

1979-01-01

The subject of radiation transport and shielding in the nuclear power industry is reviewed, and advances in the state of the art are described. These fall into the areas of computational methods, nuclear cross sections, industry practices, and standards. Computer codes and data available from the Radiation Shielding Information Center (RSIC) representing recent advances are also described

Onboard radiation shielding estimates for interplanetary manned missions

International Nuclear Information System (INIS)

Totemeier, A.; Jevremovic, T.; Hounshel, D.

2004-01-01

The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

Validation of nuclear models used in space radiation shielding applications

International Nuclear Information System (INIS)

Norman, Ryan B.; Blattnig, Steve R.

2013-01-01

A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

Early test facilities and analytic methods for radiation shielding: Proceedings

International Nuclear Information System (INIS)

Ingersoll, D.T.; Ingersoll, J.K.

1992-11-01

This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone?, a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory

Radiation shielding device

International Nuclear Information System (INIS)

Nakagawa, Takahiro; Yamagami, Makoto.

1996-01-01

A fixed shielding member made of a radiation shielding material is constituted in perpendicular to an opening formed on radiation shielding walls. The fixed shielding member has one side opened and has other side, the upper portion and the lower portion disposed in close contact with the radiation shielding walls. Movable shielding members made of a radiation shielding material are each disposed openably on both side of the fixed shielding member. The movable shielding member has a shaft as a fulcrum on one side thereof for connecting it to the radiation shielding walls. The other side has a handle attached for opening/closing the movable shielding member. Upon access of an operator, when each one of the movable shielding members is opened/closed on every time, leakage of linear or scattered radiation can be prevented. Even when both of the movable shielding members are opened simultaneously, the fixed shielding member and the movable shielding members form labyrinth to prevent leakage of linear radioactivity. (I.N.)

SP-100 GES/NAT radiation shielding systems design and development testing

International Nuclear Information System (INIS)

Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

1991-01-01

Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

Advanced methodologies of evaluating the radiation sources and ionising radiation shieldings for reducing the irradiation in nuclear field personnel

International Nuclear Information System (INIS)

Pantazi, D.; Mateescu, S.; Stanciu, M.

2003-01-01

One of the technical measures of protection against ionizing radiations is the radiation shielding. The process of implementing modern and efficient methods of evaluating the radiation shielding implies advanced calculation methods. That means using from simpler 1-D or 2-D computing codes such as MicroShield or QAD up to systems of codes such as SCALE (containing several independent modules) or the Monte Carlo multipurpose and many particles, MCNP, transport code. The main objective of this work is to present the Monte Carlo based evaluation of the dose rates from the CANDU type spent fuel all along the path of its handling up to intermediate storage. These values will be then compared with the values obtained from calculations with different computing programs. To obtain this objective two problems were approached: - establishing geometrical models according to the definition used by MCNP code so that the characteristics of CANDU type nuclear fuel are taking into account; - checking the validity of the proposed models by comparing the MCNP results with those obtained with other computing codes specific for shielding evaluation and radiation dose calculation

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding)

International Nuclear Information System (INIS)

1978-01-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976

Safety guide data on radiation shielding in a reprocessing facility

International Nuclear Information System (INIS)

Sekiguchi, Noboru; Naito, Yoshitaka

1986-04-01

In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

Proceedings of a meeting on radiation shielding and related topics

International Nuclear Information System (INIS)

1978-01-01

This is a proceedings of a meeting on radiation shielding and related topics held on Feb. 22 and 23 in 1978 at Nuclear Engineering Research Laboratory of University of Tokyo. The reports includes the following items (1) studies on neutronics with accelerators (2) radiation damage (3) shielding design (4) radiation streaming (5) shielding experiments from a point of view of radiation measurements (6) shielding benchmark experiments (7) prospects on the study of neutronics. All items are written in Japanese. (auth.)

Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

International Nuclear Information System (INIS)

1980-05-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

Radiation protection/shield design: a need for a systems approach

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods are discussed

Radiation shield analysis for a manned Mars rover

International Nuclear Information System (INIS)

Morley, N.J.; ElGenk, M.S.

1991-01-01

Radiation shielding for unmanned space missions has been extensively studied; however, designs of man-rated shields are minimal. Engle et al.'s analysis of a man-rated, multilayered shield composed of two and three cycles (a cycle consists of a tungsten and a lithium hydride layer) is the basis for the work reported in this paper. The authors present the results of a recent study of shield designs for a manned Mars rover powered by a 500-kW(thermal) nuclear reactor. A train-type rover vehicle was developed, which consists of four cars and is powered by an SP-100-type nuclear reactor heat source. The maximum permissible dose rate (MPD) from all sources is given by the National Council on Radiation Protection and Measurements as 500 mSv/yr (50 rem/yr) A 3-yr Mars mission (2-yr round trip and 1-yr stay) will deliver a 1-Sv natural radiation dose without a solar particle event, 450 mSv/yr in flight, and an additional 100 mSv on the planet surface. An anomalously large solar particle event could increase the natural radiation dose for unshielded astronauts on the Martian surface to 200 mSv. This limits the MPD to crew members from the nuclear reactor to 300 mSv

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding). [1973--1976

Energy Technology Data Exchange (ETDEWEB)

1978-01-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976.

Concrete shielding for nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

1983-01-01

The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

Concrete shielding for nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagase, Tetsuo; Saito, Tetsuo

1983-01-01

The repair works of the shielding for the nuclear ship ''Mutsu'' were completed in August, 1982. For the primary shielding, serpentine concrete was adopted as it contains a large quantity of water required for neutron shielding, and in the secondary shielding at the upper part of the reactor containment vessel, the original shielding was abolished, and the heavy concrete (high water content, high density concrete) which is effective for neutron and gamma-ray shielding was newly adopted. In this report, the design and construction using these shielding concrete are outlined. In September, 1974, Mutsu caused radiation leak during the test, and the cause was found to be the fast neutrons streaming through a gap between the reactor pressure vessel and the primary shielding. The repair works were carried out in the Sasebo Shipyard. The outline of the repair works of the shielding is described. The design condition for the shielding, the design standard for the radiation dose outside and inside the ship, the method of shielding analysis and the performance required for shielding concrete are reported. The selection of materials, the method of construction and mixing ratio, the evaluation of the soundness and properties of concrete, and the works of placing the shielding concrete are outlined. (Kako, I.)

Radiation shielding plate

International Nuclear Information System (INIS)

Kobayashi, Torakichi; Sugawara, Takeo.

1983-01-01

Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

Radiation protection and shielding standards for the 1980s

International Nuclear Information System (INIS)

Trubey, D.K.

1982-01-01

The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society

Nuclear radiation and the properties of concrete

International Nuclear Information System (INIS)

Kaplan, M.F.

1983-08-01

Concrete is used for structures in which the concrete is exposed to nuclear radiation. Exposure to nuclear radiation may affect the properties of concrete. The report mentions the types of nuclear radiation while radiation damage in concrete is discussed. Attention is also given to the effects of neutron and gamma radiation on compressive and tensile strength of concrete. Finally radiation shielding, the attenuation of nuclear radiation and the value of concrete as a shielding material is discussed

Heavy density concrete for nuclear radiation shielding and power stations: [Part]3

International Nuclear Information System (INIS)

Singha Roy, P.K.

1987-01-01

This article is the third part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. Specific considerations relevant to natural but manufactured heavy aggregates like haematite used in India are briefly discussed. They include water-cement ratio, strength versus water-cement ratio, mix design strength and aggregate grading. Some typical mix proportions in haematite concretes used in India are given. Equipment for heavy density concrete is mentioned. Quality control methods and tests for heavy density concrete are described under the heading: type and chemical composition of the rock, specific gravity and surface absorption of the aggregates, grading of aggregates, cement, batching, mixing, compressive strength, and density. Construction aspects such as form work, placement, vibration, finishing, and temperature control are discussed. Finally it is pointed out that for optimising the design and economy of heavy density concrete, it is necessary to carry out country-wide survey of suitable materials, to study their properties, suitability and effectiveness in shielding radiation. (M.G.B.)

Handbook of radiation shielding data

International Nuclear Information System (INIS)

Courtney, J.C.

1976-07-01

This handbook is a compilation of data on units, conversion factors, geometric considerations, sources of radiation, and the attenuation of photons, neutrons, and charged particles. It also includes related topics in health physics. Data are presented in tabular and graphical form with sufficient narrative for a least first-approximation solutions to a variety of problems in nuclear radiation protection. Members of the radiation shielding community contributed the information in this document from unclassified and uncopyrighted sources, as referenced

Shield design development of nuclear propulsion merchant ship

International Nuclear Information System (INIS)

Tanaka, Yoshihisa

1975-01-01

Shielding design both in Japan and abroad for nuclear propulsion merchant ships is explained, with emphasis on the various technological problems having occurred in the shield design for one-body type and separate type LWRs as conceptual design. The following matters are described: the peculiarities of the design as compared with the case of land-based nuclear reactors, problems in the design standards of shielding, the present status and development of the design methods, and the instances of the design; thereby, the trends of shielding design are disclosed. The following matters are pointed out: Importance of the optimum design, of shielding, significance of radiation streaming through large voids, activation of the secondary water in built-in type steam generators, and the need of the guides for shield design. (Mori, K.)

Study of neutron and gamma shielding by lead borate and bismuth lead borate glasses: transparent radiation shielding

International Nuclear Information System (INIS)

Singh, Vishwanath P.; Badiger, N.M.

2013-01-01

Radiation shielding for gamma and neutron is the prominent area in nuclear reactor technology, medical application, dosimetry and other industries. Shielding of these types of radiation requires an appropriate concrete with mixture of low-to-high Z elements which is an opaque medium. The transparent radiation shielding in visible light for gamma and neutron is also extremely essential in the nuclear facilities as lead window. Presently various types of lead equivalent glass oxides have been invented which are transparent as well as provide protection from radiation. In our study we have assessment of effectiveness of neutron and gamma radiation shielding of xPbO.(1-x) B 2 O 3 (x=0.15 to 0.60) and xBi 2 O 3 .(0.80-x) PbO.0.20 B 2 O 3 (x=0.10 to 0.70) transparent borate and bismuth glasses by NXCOM program. The neutron effective mass removal cross section, Σ R /ρ (cm 2 /g) of the lead, bismuth and boron oxides are given. We found invariable Σ R /ρ of various combinations of the lead borate glass for x=0.15 to 0.60 and bismuth lead borate glass for x=0.10 to 0.70. It is observed that the effective removal cross-section for fast neutron (cm -1 ) of lead borate reduces significantly whereas roughly constant for bismuth borate. The gamma mass attenuation coefficients (μ/ρ) of the glasses were also compared with possible experimental values and found comparable. High (μ/ρ) for gamma radiation of the bismuth glasses shows that it is better gamma shielding compared with lead containing glass. However lead borate glasses are better neutron shielding as the neutron removal coefficient are higher. Our investigation is very useful for nuclear reactor technology where prompt neutron of energy 17 MeV and gamma photon up to 10 MeV produced. (author)

Improvements in or relating to nuclear shields

International Nuclear Information System (INIS)

Hawkins, R.J.; Riley, K.; Powell, C.

1981-01-01

A nuclear radiation shield comprises two pieces of steel held together edge to edge by a weld, the depth of which is less than the thickness of either of the edges. As the radiaion shielding effect of the weld will be less than the steel, an insert is bolted or welded over the weld. (U.K.)

Development and production of radiation shielding window (RSW) glass: Indian scenario

International Nuclear Information System (INIS)

Phani, K.K.

2006-01-01

Nuclear energy/power and its peaceful applications play an ever increasing role in India. Irradiated nuclear fuels, irradiated structural materials from reactors, nuclear wastes and radio-isotopes emit high energy gamma radiations which are extremely health hazardous. These materials are handled remotely by manipulators inside the hot cells, which are constructed by shielding materials such as lead and concrete walls. The direct visual control of processes in the hot cells during operation demands the windows in the radiation shielding walls. These windows must provide the clear viewing but yet ensure the good protection to the working personnel from the high energy radiation

Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

Science.gov (United States)

Basyigit, Celalettin; Uysal, Volkan; Kilinçarslan, Şemsettin; Mavi, Betül; Günoǧlu, Kadir; Akkurt, Iskender; Akkaş, Ayşe

2011-12-01

The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

International Nuclear Information System (INIS)

Basyigit, Celalettin; Uysal, Volkan; Kilincarslan, Semsettin; Akkas, Ayse; Mavi, Betuel; Guenoglu, Kadir; Akkurt, Iskender

2011-01-01

The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

Radiation shielding lead shield

International Nuclear Information System (INIS)

Dei, Shoichi.

1991-01-01

The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

Durability and shielding performance of borated Ceramicrete coatings in beta and gamma radiation fields

Energy Technology Data Exchange (ETDEWEB)

Wagh, Arun S., E-mail: asw@anl.gov [Environmental Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Sayenko, S.Yu.; Dovbnya, A.N.; Shkuropatenko, V.A.; Tarasov, R.V.; Rybka, A.V.; Zakharchenko, A.A. [National Science Center, Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

2015-07-15

Highlights: • It incorporates all suggestions by the reviewers. • Explanation to each new term is provided and suitable references are given. • Sample identities have been streamlined by revising the text and the tables. • Some figures have been redrawn. - Abstract: Ceramicrete™, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate. Fillers are used to impart desired properties to the product. Ceramicrete’s tailored compositions have resulted in several commercial structural products, including corrosion- and fire-protection coatings. Their borated version, called Borobond™, has been studied for its neutron shielding capabilities and is being used in structures built for storage of nuclear materials. This investigation assesses the durability and shielding performance of borated Ceramicrete coatings when exposed to gamma and beta radiations to predict the composition needed for optimal shielding performance in a realistic nuclear radiation field. Investigations were conducted using experimental data coupled with predictive Monte Carlo computer model. The results show that it is possible to produce products for simultaneous shielding of all three types of nuclear radiations, viz., neutrons, gamma-, and beta-rays. Additionally, because sprayable Ceramicrete coatings exhibit excellent corrosion- and fire-protection characteristics on steel, this research also establishes an opportunity to produce thick coatings to enhance the shielding performance of corrosion and fire protection coatings for use in high radiation environment in nuclear industry.

Numerical simulation of a reinforced concrete shield around a nuclear reactor

International Nuclear Information System (INIS)

Mahama, Mumuni Salifu

1996-02-01

Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)

Radiation shielding estimates for manned Mars space flight

International Nuclear Information System (INIS)

Dudkin, V.E.; Kovalev, E.E.; Kolomensky, A.V.; Sakovich, V.A.; Semenov, V.F.; Demin, V.P.; Benton, E.V.

1992-01-01

In the analysis of the required radiation shielding for spacecraft during a Mars flight, the specific effects of solar activity (SA) on the intensity of galactic and solar cosmic rays were taken into consideration. Three spaceflight periods were considered: (1) maximum SA; (2) minimum SA; and (3) intermediate SA, when intensities of both galactic and solar cosmic rays are moderately high. Scenarios of spaceflights utilizing liquid-propellant rocket engines, low-and intermediate-thrust nuclear electrojet engines, and nuclear rocket engines, all of which have been designed in the Soviet Union, are reviewed. Calculations were performed on the basis of a set of standards for radiation protection approved by the U.S.S.R. State Committee for Standards. It was found that the lowest estimated mass of a Mars spacecraft, including the radiation shielding mass, obtained using a combination of a liquid propellant engine with low and intermediate thrust nuclear electrojet engines, would be 500-550 metric tons. (author)

Radiation shielding concrete

International Nuclear Information System (INIS)

Kunishima, Shigeru.

1990-01-01

The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

Radiation shielding for fission reactors

Energy Technology Data Exchange (ETDEWEB)

Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

2000-03-01

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

Radiation shielding calculations for the vista spacecraft

International Nuclear Information System (INIS)

Sahin, Suemer; Sahin, Haci Mehmet; Acir, Adem

2005-01-01

The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm 3 ) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm 3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm 3 , and the peak γ-ray heating density is ∼3.0 mW/cm 3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6 Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm 3

Development of advanced, non-toxic, synthetic radiation shielding aggregate

Energy Technology Data Exchange (ETDEWEB)

Mudgal, Manish; Chouhan, Ramesh Kumar; Verma, Sarika; Amritphale, Sudhir Sitaram; Das, Satyabrata [CSIR-Advanced Materials and Processes Research Institute, Bhopal (India); Shrivastva, Arvind [Nuclear Power Corporation of India Ltd. (NPCIL), Mumbai (India)

2018-04-01

For the first time in the world, the capability of red mud waste has been explored for the development of advanced synthetic radiation shielding aggregate. Red mud, an aluminium industry waste consists of multi component, multi elemental characteristics. In this study, red mud from two different sources have been utilized. Chemical formulation and mineralogical designing of the red mud has been done by ceramic processing using appropriate reducing agent and additives. The chemical analysis, SEM microphotographs and XRD analysis confirms the presence of multi-component, multi shielding and multi-layered phases in both the different developed advance synthetic radiation shielding aggregate. The mechanical properties, namely aggregate impact value, aggregate crushing value and aggregate abrasion value have also been evaluated and was compared with hematite ore aggregate and found to be an excellent material useful for making advanced radiation shielding concrete for the construction of nuclear power plants and other radiation installations.

Analysis of shield for the nuclear ship MUTSU

International Nuclear Information System (INIS)

Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

1975-01-01

On the nuclear ship MUTSU, a higher-than-expected level of radiation was found, with output raised to 1.4 per cent. To investigate the radiation leakage, the analysis of the shielding problem utilized a four-step sequence of PALLAS-2DCY cylindrical r-z calculations with fixed sources distributions in the core. The neutron dose contours show the importance of streaming in the gap between the reactor vessel and the primary shield. Dominant consideration of thermal insulation exclude shielding from this area resulting in an imbalance in the shielding effectiveness. The neutron dose rate at the upper part of the reactor vessel is increased by neutrons incident on the head from cavity scattering. The calculation indicates that the neutron dose rate at the top of the primary shield is 5 rem/hr at 100 per cent output. (auth.)

Evaluation of radiation-shielding properties of the composite material

International Nuclear Information System (INIS)

Pavlenko, V.I.; Chekashina, N.I.; Yastrebinskij, R.N.; Sokolenko, I.V.; Noskov, A.V.

2016-01-01

The paper presents the evaluation of radiation-shielding properties of composite materials with respect to gamma-radiation. As a binder for the synthesis of radiation-shielding composites we used lead boronsilicate glass matrix. As filler we used nanotubular chrysotile filled with lead tungstate PbWO4. It is shown that all the developed composites have good physical-mechanical characteristics, such as compressive strength, thermal stability and can be used as structural materials. On the basis of theoretical calculation we described the graphs of the gamma-quanta linear attenuation coefficient depending on the emitted energy for all investigated composites. We founded high radiation-shielding properties of all the composites on the basis of theoretical and experimental data compared to materials conventionally used in the nuclear industry - iron, concrete, etc

Light-refractory radiation shielding materials using diatomites and zeolites

International Nuclear Information System (INIS)

Murakami, Hideki

2005-01-01

It has been recently shown that diatomites and zeolites have some useful characteristics for radiation shielding materials. In this study, the availability of these materials for unexpected accidents in the nuclear sites is examined. The diatomites and zeolites, compared to existing shielding materials, have superior characteristics; low density and light weight, low in radiation-induced problem, high-heat resistance, remain unaltered by the addition of an acid except hydrofluoric acid, porous and large specific surface area, and also excellent water-absorbing property. These porous materials could also expand the shielding energy range applied and be used for fast- and thermal-neutrons, and γ ray. In addition, these materials are easy to store for long periods of time against emergency because of their natural rocks. From the examinations, it is cleared that diatomites and zeolites have excellent properties as radiation shielding materials for emergency use. (author)

Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

Energy Technology Data Exchange (ETDEWEB)

Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

2014-08-15

A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

Study on Basic Characteristics for the Development of Radiation Shielding High-Weight Concrete

Energy Technology Data Exchange (ETDEWEB)

Mun, Young Bum; Lee, Jea Hyung; Choi, Hyun Kook [Sungshin Cement CO., Sejong (Korea, Republic of); Oh, Jeong Hwan; Choi, Soo Seok [Jeju National University, Jeju (Korea, Republic of)

2016-05-15

It is planned to build a power plant more than 6 units. Although the demand of a nuclear power plant is going to increase, the attention for radiation shielding is relatively in a low level. Concrete is one of the excellent and widely used shielding materials. Since the radiation shielding of a given material is proportional to density and thickness, a high-weight concrete with high-weight aggregate which is higher than normal concrete is used for radiation shielding. However, there are a few studies and references about radiation shielding concrete. Therefore, it is required to find a high-weight aggregate. The purpose of this paper is the development of a highweight concrete to improve radiation shielding capability. The radiation shielding rate of high-weight concrete is higher than that of reference concrete. It is confirmed that the density of aggregate and the unit weight of concreate is proportional to the radiation shielding rate. In addition, the chemical composition of aggregate has also has an important effect on γ-ray shielding. Therefore, high weight aggregates of higher density are essentially required to improve radiation shielding capability. The compressive strength of a high weight concrete is better than that of reference concrete. Slump and air contents, however, are slightly increased with by-product aggregates.

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.S.

1990-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.S.

1989-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

Assessment of radiation shielding materials for protection of space crews using CR-39 plastic nuclear track detector

International Nuclear Information System (INIS)

DeWitt, J.M.; Benton, E.R.; Uchihori, Y.; Yasuda, N.; Benton, E.V.; Frank, A.L.

2009-01-01

A significant obstacle to long duration human space exploration such as the establishment of a permanent base on the surface of the Moon or a human mission to Mars is the risk posed by prolonged exposure to space radiation. In order to keep mission costs at acceptable levels while simultaneously minimizing the risk from radiation to space crew health and safety, a judicious use of optimized shielding materials will be required. We have undertaken a comprehensive study using CR-39 plastic nuclear track detector (PNTD) to characterize the radiation shielding properties of a range of materials-both common baseline materials such as Al and polyethylene, and novel multifunctional materials such as carbon composites-at heavy ion accelerators. The study consists of analyzing CR-39 PNTD exposed in front of and behind shielding targets of varying composition and at a number of depths (target thicknesses) relevant to the development and testing of materials for space radiation shielding. Most targets consist of 10 cm x 10 cm slabs of solid materials ranging in thickness from 1 to >30 g/cm 2 . Exposures have been made to beams of C, O, Ne, Si, Ar, and Fe at energies ranging from 290 MeV/amu to 1 GeV/amu at the National Institute of Radiological Sciences HIMAC and the NASA Space Radiation Laboratory (NSRL) at Brookhaven National Laboratory. Analysis of the exposed detectors yields LET spectrum, dose, and dose equivalent as functions of target depth and composition, and incident heavy ion charge, energy, and fluence. Efforts are currently underway to properly weigh and combine these results into a single quantitative estimate of a material's ability to shield space crews from the interplanetary galactic cosmic ray flux.

Radiation shielding

International Nuclear Information System (INIS)

Aitken, D.

1979-01-01

Shields for equipment in which ionising radiation is associated with high electrical gradients, for example X-ray tubes and particle accelerators, incorporate a radiation-absorbing metal, as such or as a compound, and are electrically non-conducting and can be placed in the high electrical gradient region of the equipment. Substances disclosed include dispersions of lead, tungsten, uranium or oxides of these in acrylics polyesters, PVC, ABS, polyamides, PTFE, epoxy resins, glass or ceramics. The material used may constitute an evacuable enclosure of the equipment or may be an external shield thereof. (U.K.)

Design and fabrication of radiation shielded laser ablation ICP-MS system

Energy Technology Data Exchange (ETDEWEB)

Ha, Yeong Keong; Han, Sun Ho; Park, Soon Dal; Park, Yang Soon; Jee, Kwang Yong; Kim, Won Ho

2006-09-15

In relation to high burn up and extended fuel cycle for the fuel cycle efficiency, we need to take chemical analysis of spent nuclear fuel for the integrity of nuclear fuel at high burn up. to measure the isotopic distribution of fission product in a high burn up nuclear fuel, radiation shielded laser ablation system was designed and fabricated. By probing the sample with a laser beam, micro sampling system for the mass analyzer was successfully developed. This report describes the structural design and the function of developed radiation shielded LA system. This system will be used for the analysis of isotopic distribution from core to rim of a spent nuclear fuel prepared from the hot-cell in PIE facility and/or an irradiated fuel from research reactor.

Radiation shielding cloth

International Nuclear Information System (INIS)

Ijiri, Yasuo; Fujinuma, Tadashi; Tamura, Shoji.

1989-01-01

Radiation shielding cloth having radiation shielding layers comprising a composition of inorganic powder of high specific gravity and rubber are excellentin flexibility and comfortable to put on. However, since they are heavy in the weight, operators are tired upon putting them for a long time. In view of the above, the radiation ray shielding layers are prepared by calendering sheets obtained by preliminary molding of the composition to set the variation of the thickness within a range of +15% to -0% of prescribed thickness. Since the composition of inorganic powder at high specific gravity and rubber used for radiation ray shielding comprises a great amount of inorganic powder at high specific gravity blended therein, it is generally poor in fabricability. Therefor, it is difficult to attain fine control for the sheet thickness by merely molding a composition block at once. Then, the composition is at first preliminarily molded into a sheet-like shape which is somewhat thickener than the final thickness and then finished by calendering, by which the thickness can be reduced in average as compared with conventional products while keeping the prescribed thickness and reducing the weight reduce by so much. (N.H.)

Radiation shielding curtain

International Nuclear Information System (INIS)

Winkler, N.T.

1976-01-01

A radiation shield is described in the form of a stranded curtain made up of bead-chains whose material and geometry are selected to produce a cross-sectional density that is the equivalent of 0.25 mm or more of lead and which curtain may be mounted on various radiological devices to shield against scattered radiation while offering a minimum of obstruction to the radiologist

Evaluation of nuclear data for radiation shielding by model calculations and international co-operation aspects

International Nuclear Information System (INIS)

Canetta, E.; Maino, G.; Menapace, E.

2001-01-01

The matter is reviewed, also following previous discussions at ICRS-9, concerning evaluation and related theoretical activities on nuclear data for radiation shielding within the framework of international co-operation initiatives, according to recognised needs and priorities. Both cross-section data.- for reactions induced by neutrons and photons - and nuclear structure data have been considered. In this context, main contributions and typical results are presented from theoretical and evaluation activities at the ENEA Applied Physics Division, especially concerning neutron induced reaction data up to 20 MeV and photonuclear reaction data such as photon absorption and (gamma,n) cross-sections. Relevant aspects of algebraic nuclear models and of evaporation and pre-equilibrium models are discussed. (authors)

Shielding from cosmic radiation for interplanetary missions Active and passive methods

CERN Document Server

Spillantini, P; Durante, M; Müller-Mellin, R; Reitz, G; Rossi, L; Shurshakov, V; Sorbi, M

2007-01-01

Shielding is arguably the main countermeasure for the exposure to cosmic radiation during interplanetary exploratory missions. However, shielding of cosmic rays, both of galactic or solar origin, is problematic, because of the high energy of the charged particles involved and the nuclear fragmentation occurring in shielding materials. Although computer codes can predict the shield performance in space, there is a lack of biological and physical measurements to benchmark the codes. An attractive alternative to passive, bulk material shielding is the use of electromagnetic fields to deflect the charged particles from the spacecraft target. Active shielding concepts based on electrostatic fields, plasma, or magnetic fields have been proposed in the past years, and should be revised based on recent technological improvements. To address these issues, the European Space Agency (ESA) established a Topical Team (TT) in 2002 including European experts in the field of space radiation shielding and superconducting magn...

Challenges in commercial manufacture of radiation shielding glasses

International Nuclear Information System (INIS)

Gupta, R.K.

2011-01-01

Radioactive hot-cells employ Radiation Shielding Windows (RSWs), assembled from specialty glasses, developed exclusively for nuclear industry. RSWs serve the twin purpose of direct viewing and shielding protection to the operator and use various types of radiation resistant and optically compatible glasses, such as low-density borosilicate glass; medium-density glass with up to 45% Lead and high-density glass with over 70% lead. Some glasses are Ceria-doped for enhancing their resistance threshold to radiation browning. A clear view of future requirement, capital and environmental costs could be the driving force towards bringing about changes in melting practices, encourage melting development, and enhancing collaboration. With DAE and CGCRI working in tandem, production of the entire range of RSW glasses by an Indian glass industry participant may no longer be a distant dream

Measuring space radiation shielding effectiveness

Directory of Open Access Journals (Sweden)

Bahadori Amir

2017-01-01

Full Text Available Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

Measuring space radiation shielding effectiveness

Science.gov (United States)

Bahadori, Amir; Semones, Edward; Ewert, Michael; Broyan, James; Walker, Steven

2017-09-01

Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

1982-01-01

A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

Design and Fabrication of Radiation Shielded Micro X-Ray Diffraction System

International Nuclear Information System (INIS)

Park, Yang Soon; Han, Sun Ho; Ha, Kyeong Yeong; Jee, Kwang Yong

2006-12-01

It has been observed that microstructure changes occur at the radial edge of pellet(rim) of the fuel at a high burn-up and extended fuel cycle. The thickness of a rim is some hundreds of micrometers. Despite its narrow range, a rim would affect the behaviour of nuclear fuel. To determine lattice parameter with micro-XRD at intervals as small as 30 - 50 μm in radial direction of irradiated fuel samples, a radiation shielded micro-XRD system was designed and fabricated. This report describes the concept, shielding analysis, the structural design and the fabrication of a radiation shielded glove box for micro-XRD system. This radiation shielded micro-XRD system will be used for analysis of lattice parameter change and the phase distribution at intervals as small as 30 - 50 μm in radial direction of the rim of an irradiated fuel sample and a fuel cladding

Design and Fabrication of Radiation Shielded Micro X-Ray Diffraction System

Energy Technology Data Exchange (ETDEWEB)

Park, Yang Soon; Han, Sun Ho; Ha, Kyeong Yeong; Jee, Kwang Yong

2006-12-15

It has been observed that microstructure changes occur at the radial edge of pellet(rim) of the fuel at a high burn-up and extended fuel cycle. The thickness of a rim is some hundreds of micrometers. Despite its narrow range, a rim would affect the behaviour of nuclear fuel. To determine lattice parameter with micro-XRD at intervals as small as 30 - 50 {mu}m in radial direction of irradiated fuel samples, a radiation shielded micro-XRD system was designed and fabricated. This report describes the concept, shielding analysis, the structural design and the fabrication of a radiation shielded glove box for micro-XRD system. This radiation shielded micro-XRD system will be used for analysis of lattice parameter change and the phase distribution at intervals as small as 30 - 50 {mu}m in radial direction of the rim of an irradiated fuel sample and a fuel cladding.

Radiation shielding and safety design

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Ouk; Gil, C. S.; Cho, Y. S.; Kim, D. H.; Kim, H. I.; Kim, J. W.; Lee, C. W.; Kim, K. Y.; Kim, B. H. [KAERI, Daejeon (Korea, Republic of)

2011-07-15

A benchmarking for the test facility, evaluations of the prompt radiation fields, evaluation of the induced activities in the facility, and estimation of the radiological impact on the environment were performed in this study. and the radiation safety analysis report for nuclear licensing was written based on this study. In the benchmark calculation, the neutron spectra was measured in the 20 Mev test facility and the measurements were compared with the computational results to verify the calculation system. In the evaluation of the prompt radiation fields, the shielding design for 100 MeV target rooms, evaluations of the leakage doses from the accidents and skyshine analysis were performed. The evaluation of the induced activities were performed for the coolant, inside air, structural materials, soil and ground-water. At last, the radiation safety analysis report was written based on results from these studies

Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

International Nuclear Information System (INIS)

Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

1997-01-01

Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

Radiation shielding bricks

International Nuclear Information System (INIS)

Crowe, G.J.W.

1983-01-01

A radiation shielding brick for use in building dry walls to form radiation proof enclosures and other structures is described. It is square in shape and comprises a sandwich of an inner layer of lead or similar shielding material between outer layers of plastics material, for structural stability. The ability to mechanically interlock adjacent bricks is provided by shaping the edges as cooperating external and internal V-sections. Relatively leak-free joints are ensured by enlarging the width of the inner layer in the edge region. (author)

Measuring space radiation shielding effectiveness

OpenAIRE

Bahadori Amir; Semones Edward; Ewert Michael; Broyan James; Walker Steven

2017-01-01

Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles ...

Preliminary radiation criteria and nuclear analysis for ETF

International Nuclear Information System (INIS)

Engholm, B.A.

1980-09-01

Preliminary biological and materials radiation dose criteria for the Engineering Test Facility are described and tabulated. In keeping with the ETF Mission Statement, a key biological dose criterion is a 24-hour shutdown dose rate of 2 mrem/hr on the surface of the outboard bulk shield. Materials dose criteria, which primarily govern the inboard shield design, include 10 9 rads exposure limit to epoxy insulation, 3 x 10 -4 dpa damage to the TF coil copper stabilizer, and a total nuclear heating rate of 5 kW in the inboard TF coils. Nuclear analysis performed during FY 80 was directed primarily at the inboard and outboard bulk shielding, and at radiation streaming in the neutral beam drift ducts. Inboard and outboard shield thicknesses to achieve the biological and materials radiation criteria are 75 cm inboard and 125 cm outboard, the configuration consisting of alternating layers of stainless steel and borated water. The outboard shield also includes a 5 cm layer of lead. NBI duct streaming analyses performed by ORNL and LASL will play a key role in the design of the duct and NBI shielding in FY 81. The NBI aluminum cryopanel nuclear heating rate during the heating cycle is about 1 milliwatt/cm 3 , which is far less than the permissible limit

Radiation shielding properties of high performance concrete reinforced with basalt fibers infused with natural and enriched boron

Energy Technology Data Exchange (ETDEWEB)

Zorla, Eyüp; Ipbüker, Cagatay [University of Tartu, Institute of Physics (Estonia); Biland, Alex [US Basalt Corp., Houston (United States); Kiisk, Madis [University of Tartu, Institute of Physics (Estonia); Kovaljov, Sergei [OÜ Basaltest, Tartu (Estonia); Tkaczyk, Alan H. [University of Tartu, Institute of Physics (Estonia); Gulik, Volodymyr, E-mail: volodymyr.gulik@gmail.com [Institute for Safety Problems of Nuclear Power Plants, Lysogirska 12, of. 201, 03028 Kyiv (Ukraine)

2017-03-15

Highlights: • Basalt fiber infused with natural and enriched boron in varying proportions. • Gamma-ray attenuation remains stable with addition of basalt-boron fiber. • Improvement in neutron shielding for nuclear facilities producing fast fission spectrum. • Basalt-boron fiber could decrease the shielding thickness in thermal spectrum reactors. - Abstract: The importance of radiation shielding is increasing in parallel with the expansion of the application areas of nuclear technologies. This study investigates the radiation shielding properties of two types of high strength concrete reinforced with basalt fibers infused with 12–20% boron oxide, containing varying fractions of natural and enriched boron. The gamma-ray shielding characteristics are analyzed with the help of the WinXCom, whereas the neutron shielding characteristics are modeled and computed by Monte Carlo Serpent code. For gamma-ray shielding, the attenuation coefficients of the studied samples do not display any significant variation due to the addition of basalt-boron fibers at any mixing proportion. For neutron shielding, the addition of basalt-boron fiber has negligible effects in the case of very fast neutrons (14 MeV), but it could considerably improve the neutron shielding of concrete for nuclear facilities producing a fast fission spectrum (e.g. with reactors as BN-800, FBTR) and thermal neutron spectrum (Light Water Reactors (LWR)). It was also found that basalt-boron fiber could decrease the thickness of radiation shielding material in thermal spectrum reactors.

Radiation shielding properties of high performance concrete reinforced with basalt fibers infused with natural and enriched boron

International Nuclear Information System (INIS)

Zorla, Eyüp; Ipbüker, Cagatay; Biland, Alex; Kiisk, Madis; Kovaljov, Sergei; Tkaczyk, Alan H.; Gulik, Volodymyr

2017-01-01

Highlights: • Basalt fiber infused with natural and enriched boron in varying proportions. • Gamma-ray attenuation remains stable with addition of basalt-boron fiber. • Improvement in neutron shielding for nuclear facilities producing fast fission spectrum. • Basalt-boron fiber could decrease the shielding thickness in thermal spectrum reactors. - Abstract: The importance of radiation shielding is increasing in parallel with the expansion of the application areas of nuclear technologies. This study investigates the radiation shielding properties of two types of high strength concrete reinforced with basalt fibers infused with 12–20% boron oxide, containing varying fractions of natural and enriched boron. The gamma-ray shielding characteristics are analyzed with the help of the WinXCom, whereas the neutron shielding characteristics are modeled and computed by Monte Carlo Serpent code. For gamma-ray shielding, the attenuation coefficients of the studied samples do not display any significant variation due to the addition of basalt-boron fibers at any mixing proportion. For neutron shielding, the addition of basalt-boron fiber has negligible effects in the case of very fast neutrons (14 MeV), but it could considerably improve the neutron shielding of concrete for nuclear facilities producing a fast fission spectrum (e.g. with reactors as BN-800, FBTR) and thermal neutron spectrum (Light Water Reactors (LWR)). It was also found that basalt-boron fiber could decrease the thickness of radiation shielding material in thermal spectrum reactors.

Study of Radiation Shielding Analysis for Low-Intermediate Level Waste Transport Ship

Energy Technology Data Exchange (ETDEWEB)

Kim, Dohyung; Lee, Unjang; Song, Yangsoo; Kim, Sukhoon; Ko, Jaehoon [Korea Nuclear Engineering and Service Corporation, Seoul (Korea, Republic of)

2007-07-01

In Korea, it is planed to transport Low-Intermediate Level Radioactive Waste (LILW) from each nuclear power plant site to Kyongju LILW repository after 2009. Transport through the sea using ship is one of the most prospective ways of LILW transport for current situation in Korea. There are domestic and international regulations for radiation dose limit for radioactive material transport. In this article, radiation shielding analysis for LILW transport ship is performed using 3-D computer simulation code, MCNP. As a result, the thickness and materials for radiation shielding walls next to cargo in the LILW transport ship are determined.

Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

International Nuclear Information System (INIS)

Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

1984-01-01

Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

Minimizing exposure in nuclear medicine through optimum use of shielding devices

International Nuclear Information System (INIS)

Rutherford, B.L.; King, S.H.; Erdman, M.C.; Miller, K.L.

1991-01-01

Exposure to radiation from nuclear medicine nuclides can be minimized through the use of various shielding devices. This paper reviews the dose reductions achieved through use of various syringe shields, lead aprons, leaded gloves, and several types of eyeglass lenses for 67 Ga, 99m Tc, 131 I and 201 Tl. The authors have found that combination of devices can best provide for minimizing doses

Radiation shielding material

International Nuclear Information System (INIS)

Kawakubo, Takamasa; Yamada, Fumiyuki; Nakazato, Kenjiro.

1976-01-01

Purpose: To provide a material, which is used for printing a samples name and date on an X-ray photographic film at the same time an X-ray radiography. Constitution: A radiation shielding material of a large mass absorption coefficient such as lead oxide, barium oxide, barium sulfate, etc. is added to a solution of a radiation permeable substance capable of imparting cold plastic fluidity (such as microcrystalline wax, paraffin, low molecular polyethylene, polyvinyl chloride, etc.). The resultant system is agitated and then cooled, and thereafter it is press fitted to or bonded to a base in the form of a film of a predetermined thickness. This radiation shielding layer is scraped off by using a writing tool to enter information to be printed in a photographic film, and then it is laid over the film and exposed to X-radiation to thereby print the information on the film. (Seki, T.)

Application of the personnel photographic monitoring method to determine equivalent radiation dose beyond proton accelerator shielding

International Nuclear Information System (INIS)

Gel'fand, E.K.; Komochkov, M.M.; Man'ko, B.V.; Salatskaya, M.I.; Sychev, B.S.

1980-01-01

Calculations of regularities to form radiation dose beyond proton accelerator shielding are carried out. Numerical data on photographic monitoring dosemeter in radiation fields investigated are obtained. It was shown how to determine the total equivalent dose of radiation fields beyond proton accelerator shielding by means of the photographic monitoring method by introduction into the procedure of considering nuclear emulsions of division of particle tracks into the black and grey ones. A comparison of experimental and calculational data has shown the applicability of the used calculation method for modelling dose radiation characteristics beyond proton accelerator shielding [ru

Radiation Shielding Materials and Containers Incorporating Same

Energy Technology Data Exchange (ETDEWEB)

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Radiation shielding glass

International Nuclear Information System (INIS)

Kido, Kazuhiro; Ueda, Hajime.

1997-01-01

It was found that a glass composition comprising, as essential ingredients, SiO 2 , PbO, Gd 2 O 3 and alkali metal oxides can provide a shielding performance against electromagnetic waves, charged particles and neutrons. The present invention provides radiation shielding glass containing at least from 16 to 46wt% of SiO 2 , from 47 to 75wt% of PbO, from 1 to 10wt% of Gd 2 O 3 , from 0 to 3wt% of Li 2 O, from 0 to 7wt% of Na 2 O, from 0 to 7wt% of K 2 O provided that Li 2 O + Na 2 O + K 2 O is from 1 to 10wt%, B 2 O 3 is from 0 to 10wt%, CeO 2 is from 0 to 3wt%, As 2 O 3 is from 0 to 1wt% and Sb 2 O 3 is from 0 to 1wt%. Since the glass can shield electromagnetic waves, charged particles and neutrons simultaneously, radiation shielding windows can be designed and manufactured at a reduced thickness and by less constitutional numbers in a circumstance where they are present altogether. (T.M.)

Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

International Nuclear Information System (INIS)

Abulfaraj, W.H.; Kamal, S.M.

1994-01-01

The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concrete shielding. Multiattribute utility theory is selected to accommodate decision maker's preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Illmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy weight heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Illmenite Serpentine concrete. (Author)

Nuclear shields

International Nuclear Information System (INIS)

Linares, R.C.; Nienart, L.F.; Toelcke, G.A.

1976-01-01

A process is described for preparing melt-processable nuclear shielding compositions from chloro-fluoro substituted ethylene polymers, particularly PCTFE and E-CTFE, containing 1 to 75 percent by weight of a gadolinium compound. 13 claims, no drawings

Hybrid Active-Passive Radiation Shielding System

Data.gov (United States)

National Aeronautics and Space Administration — A radiation shielding system is proposed that integrates active magnetic fields with passive shielding materials. The objective is to increase the shielding...

Radiation shielding activities at IDOM

Energy Technology Data Exchange (ETDEWEB)

Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora, E-mail: cesar.hueso@idom.com [IDOM, Consulting, Engineering and Architecture, S.A.U, Vizcaya (Spain)

2017-07-01

When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

Radiation shielding activities at IDOM

International Nuclear Information System (INIS)

Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora

2017-01-01

When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

Study and application of Dot 3.5 computer code in radiation shielding problems

International Nuclear Information System (INIS)

Otto, A.C.; Mendonca, A.G.; Maiorino, J.R.

1983-01-01

The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.) [pt

Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures.

Science.gov (United States)

Dickson, E D; Hamby, D M

2014-03-01

The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building's radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology.

Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M

2014-01-01

The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building’s radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology. (paper)

Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

2001-08-15

Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Science.gov (United States)

Remec, Igor; Rosseel, Thomas M.; Field, Kevin G.; Pape, Yann Le

2017-09-01

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958

International Nuclear Information System (INIS)

Goebelbecker, Hans-Juergen

2008-01-01

Description: The papers of the European Atomic Energy Society Symposium VI-58 on radiation shielding (ICRS1) held at Caius College, Cambridge England from 26 to 29 August 1958 are collected here for the first time in electronic form. This symposium was organised in connection with the Second Atoms for Peace Conference held in Geneva Held in Geneva from 1 to 13 September 1958. The Topics discussed covered gamma rays and neutron radiation; the Methods discussed were analytical approaches, semi-empirical Methods, simple computer codes, Monte Carlo method. Little quality nuclear data for shielding calculations was available and the presentations would concentrate on removal cross-sections and build-up factors. Experimental techniques in support to estimate the effective shielding properties of materials were discussed such as general experimental shielding techniques and experiments on neutron attenuation in different materials and on concrete as shield. Foil detectors for spectra measurements and determination of dose rates were mainly used. The typical issues addressed were gamma-heating, gamma spectra, neutron induced gammas, fission products gamma spectra, skyshine radiation and neutron ducts - streaming. Most participants were researchers from the naval and aeronautics sector

Highly heat removing radiation shielding material

International Nuclear Information System (INIS)

Asano, Norio; Hozumi, Masahiro.

1990-01-01

Organic materials, inorganic materials or metals having excellent radiation shielding performance are impregnated into expanded metal materials, such as Al, Cu or Mg, having high heat conductivity. Further, the porosity of the expanded metals and combination of the expanded metals and the materials to be impregnated are changed depending on the purpose. Further, a plurality of shielding materials are impregnated into the expanded metal of the same kind, to constitute shielding materials. In such shielding materials, impregnated materials provide shielding performance against radiation rays such as neutrons and gamma rays, the expanded metals provide heat removing performance respectively and they act as shielding materials having heat removing performance as a whole. Accordingly, problems of non-informity and discontinuity in the prior art can be dissolved be provide materials having flexibility in view of fabrication work. (T.M.)

Irrigoscopy - irrigography method, dosimetry and radiation shielding

International Nuclear Information System (INIS)

Zubanov, Z.; Kolarevic, G.

1999-01-01

Use of patient's radiation shielding during radiology diagnostic procedures in our country is insufficiently represent, so patients needlessly receive very high entrance skin doses in body areas which are not in direct x-ray beam. During irrigoscopy, patient's radiation shielding is very complex problem, because of the organs position. In the future that problem must be solved. We hope that some of our suggestions about patient's radiation shielding during irrigoscopy, can be a small step in that way. (author)

The construction of radiation shielding for baby ebm

International Nuclear Information System (INIS)

Mohd Rizal Md Chulan; Leo Kwee Wah; Lee Chee Huei; Muhamad Zahidee Taat; Fadzlie Nordin; Abu Bakar Mhd Ghazali; Mohd Yusof Ali; Mohd Rizal Mamat Ibrahim; Syed Nasaruddin Syed Idris; Mahmud Hamid; Mohd Khairi Mohd Said

2005-01-01

The construction of radiation shielding for electron beam machine, Baby EBM is necessary for prevention from x-ray (Bremstrahlung) that produced when electron bombarded the target material. The strength of produced x-ray is depending on electron energy and the atomic number of target material. In the construction process of radiation shielding, a few aspects need to be considered such as shielding material and its thickness to be used, mainframe for radiation shielding and the way fabrication to be done. In this project, the thickness of radiation shielding is calculated manually following the NCRP 51 guidelines whereas for frame design, shielding walls and fabrication is considered that the accelerator devices (accelerating tube, focusing device and neck) is vertically and the whole weight of Baby EBM. From the calculations, the thickness and the material for radiation shielding is to be used are 6mm lead. This radiation shielding has been tested (using the parameters that have been considered) to know the leak of radiation (at all surfaces) and direct radiation below 5 cm from the window. The value of high voltage that applied at accelerating tube is 80 kV and the voltage, current supply at electron gun is 3.0 V, 7.1 A respectively. The result of the testing found that dose rate under the window foil is more than 2000 mSv/hr and at all shielding surfaces are less than 0.5 mSv/hr, which is background reading and this is acceptable as compared to the theoretical calculation. The measurement was done using a survey meter typed Ludlum-model 3. (Author)

The Radiation Safety Information Computational Center (RSICC): A Resource for Nuclear Science Applications

International Nuclear Information System (INIS)

Kirk, Bernadette Lugue

2009-01-01

The Radiation Safety Information Computational Center (RSICC) has been in existence since 1963. RSICC collects, organizes, evaluates and disseminates technical information (software and nuclear data) involving the transport of neutral and charged particle radiation, and shielding and protection from the radiation associated with: nuclear weapons and materials, fission and fusion reactors, outer space, accelerators, medical facilities, and nuclear waste management. RSICC serves over 12,000 scientists and engineers from about 100 countries. An important activity of RSICC is its participation in international efforts on computational and experimental benchmarks. An example is the Shielding Integral Benchmarks Archival Database (SINBAD), which includes shielding benchmarks for fission, fusion and accelerators. RSICC is funded by the United States Department of Energy, Department of Homeland Security and Nuclear Regulatory Commission.

Radiation shielding fiber and its manufacturing method

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Koji; Ono, Hiroshi.

1988-08-17

Purpose: To manufacture radiation shielding fibers of excellent shielding effects. Method: Fibers containing more than 1 mmol/g of carboxyl groups are bonded with heavy metals, or they are impregnated with an aqueous solution containing water-soluble heavy metal salts dissolved therein. Fibers as the substrate may be any of forms such as short fibers, long fibers, fiber tows, webs, threads, knitting or woven products, non-woven fabrics, etc. It is however necessary that fibers contain more than 1 mmol/g, preferably, from 2 to 7 mmol/g of carboxylic groups. Since heavy metals having radiation shielding performance are bonded to the outer layer of the fibers and the inherent performance of the fibers per se is possessed, excellent radiation shielding performance can be obtained, as well as they can be applied with spinning, knitting or weaving, stitching, etc. thus can be used for secondary fiber products such as clothings, caps, masks, curtains, carpets, cloths, etc. for use in radiation shieldings. (Kamimura, M.).

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

Energy Technology Data Exchange (ETDEWEB)

SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

2000-07-01

Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

Boron filled siloxane polymers for radiation shielding

Science.gov (United States)

Labouriau, Andrea; Robison, Tom; Shonrock, Clinton; Simmonds, Steve; Cox, Brad; Pacheco, Adam; Cady, Carl

2018-03-01

The purpose of the present work was to evaluate changes to structure-property relationships of 10B filled siloxane-based polymers when exposed to nuclear reactor radiation. Highly filled polysiloxanes were synthesized with the intent of fabricating materials that could shield high neutron fluences. The newly formulated materials consisted of cross-linked poly-diphenyl-methylsiloxane filled with natural boron and carbon nanofibers. This polymer was chosen because of its good thermal and chemical stabilities, as well as resistance to ionizing radiation thanks to the presence of aromatic groups in the siloxane backbone. Highly isotopically enriched 10B filler was used to provide an efficient neutron radiation shield, and carbon nanofibers were added to improve mechanical strength. This novel polymeric material was exposed in the Annular Core Research Reactor (ACRR) at Sandia National Labs to five different neutron/gamma fluxes consisting of very high neutron fluences within very short time periods. Thermocouples placed on the specimens recorded in-situ temperature changes during radiation exposure, which agreed well with those obtained from our MCNP simulations. Changes in the microstructural, thermal, chemical, and mechanical properties were evaluated by SEM, DSC, TGA, FT-IR NMR, solvent swelling, and uniaxial compressive load measurements. Our results demonstrate that these newly formulated materials are well-suitable to be used in applications that require exposure to different types of ionizing conditions that take place simultaneously.

Work for radiation shielding concrete in large-scaled radiation facilities

International Nuclear Information System (INIS)

Konomi, Shinzo; Sato, Shoni; Otake, Takao.

1980-01-01

This paper reports the radiation shielding concrete work in the construction of radiation laboratory facilities of Electrotechnical Laboratory, a Japanese Government agency for the research and development of electronic technology. The radiation shielding walls of the facilities are made of ordinary concrete, heavy weight concrete and raw iron ore. This paper particularly relates the use of ordinary concrete which constitutes the majority of such concretes. The concrete mix was determined so as to increase its specific gravity for better shielding effect, to improve mass concrete effect and to advance good workability. The tendency of the concrete to decrease its specific gravity and the temperature variations were also made on how to place concrete to secure good shielding effect and uniform quality. (author)

Physics contributions to radiation protection in nuclear power plants

International Nuclear Information System (INIS)

Krueger, F.W.

1980-01-01

Physical research and physical methods can essentially contribute to radiation protection in nuclear power plants. With their aid, properties of radiation sources can be determined, and calculations of radiation shields can be performed. In the present paper, such tasks are analyzed, the state of the art of their solution is evaluated, and trends of further work are shown. Focal points of the present study are the calculation of properties of radiation sources outside the reactor (fission products, activated corrosion products, decontamination facilities for contaminated media), exact and engineering methods for calculating radiation fields also in inhomogeneous shields, and classification of concretes for gamma-ray shielding. Objectives, possibilities, and problems of standardization of such activities are discussed. (author)

Investigation of the use of Galena concrete in electromagnetic radiation shielding

International Nuclear Information System (INIS)

Egwuonwu, G. N.; Bukar, P. H.; Avaa, A.

2011-01-01

Galena samples, collected from Ishiagu, south-eastern Nigeria, were used to make high density concretes for experimental radiation shielding. The concretes were molded into cylindrical tablets of various densities and volumes in order to ascertain their attenuation capability to some electromagnetic radiations. Blue visible light and gamma-ray sourced from cobalt-60, were transmitted through the concretes and detected with the aid of Op-Amp and digital Geiger-Muller Counter respectively. The absorption coefficients of the samples of thicknesses in the range of 1.00 - 5.00 cm were determined. Results show that for a typical galena concrete of average density 2.33gcm -3 , the absorption coefficient is about 1.186 cm -1 for the blue light and 0.495cm -1 for gamma-ray. For this density, 4.45cm of the galena concrete reduces the gamma-ray intensity by 90% and its half value layer thickness is 1.40cm. The investigation however, suggests the shielding properties of the galena sourced from Ishiagu. A database of shielding strength for the in situ galena was established hence, can serve as suitable platform for quality and quantity control in radiation shielding technology in radiotherapy treatment rooms and nuclear reactors.

Radiation shielding for TFTR DT diagnostics

International Nuclear Information System (INIS)

Ku, L.P.; Johnson, D.W.; Liew, S.L.

1994-01-01

The authors illustrate the designs of radiation shielding for the TFTR DT diagnostics using the ACX and TVTS systems as specific examples. The main emphasis here is on the radiation transport analyses carried out in support of the designs. Initial results from the DT operation indicate that the diagnostics have been functioning as anticipated and the shielding designs are satisfactory. The experience accumulated in the shielding design for the TFTR DT diagnostics should be useful and applicable to future devices, such as TPX and ITER, where many similar diagnostic systems are expected to be used

Accelerator shielding benchmark problems

International Nuclear Information System (INIS)

Hirayama, H.; Ban, S.; Nakamura, T.

1993-01-01

Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

Evaluation of syringe shield effectiveness in handling radiopharmaceuticals

Directory of Open Access Journals (Sweden)

Cho Yong-In

2015-01-01

Full Text Available 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.

Self Shielding in Nuclear Fissile Assay Using LSDS

International Nuclear Information System (INIS)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

2012-01-01

The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center. Volume 7

Energy Technology Data Exchange (ETDEWEB)

Trubey, D.K.; Roussin, R.W.; Gustin, A.B.

1983-08-01

An indexed bibliography of open literature selected by the Radiation Shielding Information Center since the previous volume was published in 1980 is presented in the area of radiation transport and shielding against radiation from nuclear reactors (fission and fusion), x-ray machines, radioisotopes, nuclear weapons (including fallout, and low energy accelerators (e.g., neutron generators). The bibliography was typeset from computer files constituting the RSIC Storage and Retrieval Information System. In addition to lists of literature titles by subject categories (accessions 6201-10156), an author index is given. Most of the literature selected for Volume VII was published in the years 1977 to 1981.

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center. Volume 7

International Nuclear Information System (INIS)

Trubey, D.K.; Roussin, R.W.; Gustin, A.B.

1983-08-01

An indexed bibliography of open literature selected by the Radiation Shielding Information Center since the previous volume was published in 1980 is presented in the area of radiation transport and shielding against radiation from nuclear reactors (fission and fusion), x-ray machines, radioisotopes, nuclear weapons (including fallout, and low energy accelerators (e.g., neutron generators). The bibliography was typeset from computer files constituting the RSIC Storage and Retrieval Information System. In addition to lists of literature titles by subject categories (accessions 6201-10156), an author index is given. Most of the literature selected for Volume VII was published in the years 1977 to 1981

Radiation shielding material

International Nuclear Information System (INIS)

Matsumoto, Akio; Isobe, Eiji.

1976-01-01

Purpose: To increase the shielding capacity of the radiation shielding material having an abundant flexibility. Constitution: A mat consisting of a lead or lead alloy fibrous material is covered with a cloth, and the two are made integral by sewing in a kilted fashion by using a yarn. Thereafter, the system is covered with a gas-tight film or sheet. The shielding material obtained in this way has, in addition to the above merits, advantages in that (1) it is free from restoration due to elasticity so that it can readily seal contaminants, (2) it can be used in a state consisting of a number of overlapped layers, (3) it fits the shoulder well and is readily portable and (4) it permits attachment of fasteners or the like. (Ikeda, J.)

Radiation Shielding Properties of Some Marbles in Turkey

International Nuclear Information System (INIS)

Guenoglu, K.; Akkurt, I.

2011-01-01

Especially after development of technology, radiation started to be used in a large fields such as medicine, industry and energy. Using radiation in those fields bring hazardous effect of radiation into humancell. Thus radiation protection becomes important in physics. Although there are three ways for radiation protection, shielding of the radiation is the most commonly used method. Natural Stones such as marble is used as construction material especially in critical building and thus its radiation shielding capability should be determined.In this study, gamma ray shielding properties of some different types of marble mined in Turkey, have been measured using a NaI(Tl) scintillator detector. The measured results were also compared with the theoretical calculations XCOM.

Shields for nuclear reactors

International Nuclear Information System (INIS)

Aspden, G.J.

1984-01-01

The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

Normalization of shielding structure quality and the method of its studying

International Nuclear Information System (INIS)

Bychkov, Ya.A.; Lavdanskij, P.A.

1987-01-01

Method for evaluation of nuclear facility radiation shield quality is suggested. Indexes of shielding structure radiation efficiency and face efficiency are used as the shielding structure quality indexes. The first index is connected with radiation dose rate during personnel irradiation behind the shield, and the second one - with the stresses in shielding structure introduction of the indexes presented allows to evaluate objectively the quality of nuclear facility shielding structure quality design construction and operation and to economize labour and material resources

Improved Metal-Polymeric Laminate Radiation Shielding, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — In this proposed Phase I program, a multifunctional lightweight radiation shield composite will be developed and fabricated. This structural radiation shielding will...

Radiation shielding properties of barite coated fabric by computer programme

Energy Technology Data Exchange (ETDEWEB)

Akarslan, F.; Molla, T. [Suleyman Demirel University, Engineering Fac. Textile Dep., Isparta (Turkey); Üncü, I. S. [Suleyman Demirel University, Technological Fac. Electrical-Electronic Eng. Dep., Isparta (Turkey); Kılıncarslan, S., E-mail: seref@tef.sdu.edu.tr [Suleyman Demirel University, Engineering Fac. Civil Eng. Dep., Isparta (Turkey); Akkurt, I. [Suleyman Demirel University, Art and Science Fac., Physics Dep., Isparta (Turkey)

2015-03-30

With the development of technology radiation started to be used in variety of different fields. As the radiation is hazardous for human health, it is important to keep radiation dose as low as possible. This is done mainly using shielding materials. Barite is one of the important materials in this purpose. As the barite is not used directly it can be used in some other materials such as fabric. For this purposes barite has been coated on fabric in order to improve radiation shielding properties of fabric. Determination of radiation shielding properties of coated fabric has been done by using computer program written C# language. With this program the images obtained from digital Rontgen films is used to determine radiation shielding properties in terms of image processing numerical values. Those values define radiation shielding and in this way the coated barite effect on radiation shielding properties of fabric has been obtained.

A study on the effect of crack in concrete structure in the point of radiation shielding

International Nuclear Information System (INIS)

Lee, Chang-Min; Lee, Yoon-Hee; Lee, Kun-Jai; Cho, Cheon-Hyung; Choi, Byung-Il; Lee, Heung-Young

2005-01-01

The saturation of South Korea's at-reactor (AR) spent fuel storage pools has created a necessity for additional spent fuel storage capacity. Because the South Korean government has a plan to increase the number of nuclear power plants to 27 units by 2016, the increase of spent nuclear fuel generation will be accelerated. Because there is no concrete plan for spent unclear fuel permanent disposal, the Korea hydraulic nuclear power company is planning to construct dry storage facility. Spent nuclear fuel from CANDU type nuclear power plant will be stored in MACSTOR-400 composed by reinforced concrete. Because it is new model, it has to be licensed. Life time estimation is needed for licensing. Deterioration of reinforced concrete structure is currently of great concern for life time estimation. The most significant form of deterioration is reinforcement corrosion that gives rise to crack the concrete structure. In this study, in order to estimate the life time of MACSTOR, the tendency of crack creation, propagation and the effect of crack in concrete structure against radiation shielding are investigated. Crack creation and propagation depends on concrete cover thickness and c/d ratio. The surface dose rate at the concrete shield in MACSTOR is simulated by MCNP code about several cases. Generally in the case of point source, surface dose rate depends on shape, width and length of crack. In the case of MACSTOR-400, It is estimated that crack is not dominant factor in the point of radiation shielding in less than 0.4mm of crack width. Above results will be helpful to estimate the life time of concrete structure as radiation shield

Nuclear characteristics of epoxy resin as a space environment neutron shielding

Energy Technology Data Exchange (ETDEWEB)

Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School

2017-03-15

In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.

Multilayer radiation shield

Science.gov (United States)

Urbahn, John Arthur; Laskaris, Evangelos Trifon

2009-06-16

A power generation system including: a generator including a rotor including a superconductive rotor coil coupled to a rotatable shaft; a first prime mover drivingly coupled to the rotatable shaft; and a thermal radiation shield, partially surrounding the rotor coil, including at least a first sheet and a second sheet spaced apart from the first sheet by centripetal force produced by the rotatable shaft. A thermal radiation shield for a generator including a rotor including a super-conductive rotor coil including: a first sheet having at least one surface formed from a low emissivity material; and at least one additional sheet having at least one surface formed from a low emissivity material spaced apart from the first sheet by centripetal force produced by the rotatable shaft, wherein each successive sheet is an incrementally greater circumferential arc length and wherein the centripetal force shapes the sheets into a substantially catenary shape.

Radiation Shielding Properties of Some Marbles in Turkey

Science.gov (United States)

Günoǧlu, K.; Akkurt, I.

2011-12-01

Especially after development of technology, radiation started to be used in a large fields such as medicine, industry and energy. Using radiation in those fields bring hazordous effect of radition into humancell. Thus radiation protection becomes important in physics. Although there are three ways for radiation protection, shielding of the radiation is the most commonly used method. Natural Stones such as marble is used as construction material especially in critical building and thus its radiation shielding capability should be determined. In this study, gamma ray shielding properties of some different types of marble mined in Turkey, have been measured using a NaI(Tl) scintillator detector. The measured results were also compared with the theoretical calculations XCOM.

Technical products for radiation shielding. Shield assembled from lead blocks for radiation protection. General technical requirements

International Nuclear Information System (INIS)

1981-01-01

The object of this standard description is the general technological requirements of 50 and 100 mm thick radiation protection shields assembled from lead blocks. The standard contains the definitions, types, parameters and dimensions of shields, their technical and acceptance criteria with testing methods, tagging, packaging, transportation and storage requirements, producer's liability. Some illustrated assembling examples, preferred parameters and dosimetry methods for shield inspection are given. (R.P.)

Development of silicone rubber-type neutron shielding material

International Nuclear Information System (INIS)

Do, Jae Bum; Cho, Soo Hang; Kim, Ik Soo; Oh, Seung Chul; Hong, Soon Seok; Noh, Sung Ki; Jeong, Duk Yeon.

1997-06-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed silicone rubber based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. (author). 16 tabs., 17 figs., 25 refs

Radiation shielding of the main injector

International Nuclear Information System (INIS)

Bhat, C.M.; Martin, P.S.

1995-05-01

The radiation shielding in the Fermilab Main Injector (FMI) complex has been carried out by adopting a number of prescribed stringent guidelines established by a previous safety analysis. Determination of the required amount of radiation shielding at various locations of the FMI has been done using Monte Carlo computations. A three dimensional ray tracing code as well as a code based upon empirical observations have been employed in certain cases

Radiation shielding wall structure

International Nuclear Information System (INIS)

Nishimura, Yoshitaka; Oka, Shinji; Kan, Toshihiko; Misato, Takeshi.

1990-01-01

A space between a pair of vertical steel plates laterally disposed in parallel at an optional distance has a structure of a plurality of vertically extending tranks partitioned laterally by vertically placed steel plates. Then, cements are grouted to the tranks. Strip-like steel plates each having a thickness greater than the gap between the each of the vertically placed steel plates and the cement are bonded each at the surface for each of the vertically placed steel plates opposing to the cements. A protrusion of a strip width having radiation shielding performance substantially identical with that by the thickness of the cement is disposed in the strip-like steel plates. With such a constitution, a safety radiation shielding wall structure with no worry of radiation intrusion to gaps, if formed, between the steel plates and the grouted cements due to shrinkage of the cements. (I.N.)

Verification of radiation exposure using lead shields

International Nuclear Information System (INIS)

Hayashida, Keiichi; Yamamoto, Kenyu; Azuma, Masami

2016-01-01

A long time use of radiation during IVR (intervention radiology) treatment leads up to an increased exposure on IVR operator. In order to prepare good environment for the operator to work without worry about exposure, the authors examined exposure reduction with the shields attached to the angiography instrument, i. e. lead curtain and lead glass. In this study, the lumber spine phantom was radiated using the instrument and the radiation leaked outside with and without shields was measured by the ionization chamber type survey meter. The meter was placed at the position which was considered to be that for IVR operator, and changed vertically 20-100 cm above X-ray focus by 10 cm interval. The radiation at the position of 80 cm above X-ray focus was maximum without shield and was hardly reduced with lead curtain. However, it was reduced with lead curtain plus lead glass. Similar reduction effects were observed at the position of 90-100 cm above X-ray focus. On the other hand, the radiation at the position of 70 cm above X-ray focus was not reduced with either shield, because that position corresponded to the gap between lead curtain and lead glass. The radiation at the position of 20-60 cm above X-ray focus was reduced with lead curtain, even if without lead glass. These results show that lead curtain and lead glass attached to the instrument can reduce the radiation exposure on IVR operator. Using these shields is considered to be one of good means for IVR operator to work safely. (author)

Shielded cells transfer automation

International Nuclear Information System (INIS)

Fisher, J.J.

1984-01-01

Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

Thick Galactic Cosmic Radiation Shielding Using Atmospheric Data

Science.gov (United States)

Youngquist, Robert C.; Nurge, Mark A.; Starr, Stanley O.; Koontz, Steven L.

2013-01-01

NASA is concerned with protecting astronauts from the effects of galactic cosmic radiation and has expended substantial effort in the development of computer models to predict the shielding obtained from various materials. However, these models were only developed for shields up to about 120 g!cm2 in thickness and have predicted that shields of this thickness are insufficient to provide adequate protection for extended deep space flights. Consequently, effort is underway to extend the range of these models to thicker shields and experimental data is required to help confirm the resulting code. In this paper empirically obtained effective dose measurements from aircraft flights in the atmosphere are used to obtain the radiation shielding function of the earth's atmosphere, a very thick shield. Obtaining this result required solving an inverse problem and the method for solving it is presented. The results are shown to be in agreement with current code in the ranges where they overlap. These results are then checked and used to predict the radiation dosage under thick shields such as planetary regolith and the atmosphere of Venus.

A study on the apron shielding ratio according to electromagnetic radiation energy

Energy Technology Data Exchange (ETDEWEB)

Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh [Dept. of Nuclear Medicine, Dongnam Institute of Radiological and Medical Sciences Cancer Center, Busan (Korea, Republic of)

2014-12-15

The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding.

A study on the apron shielding ratio according to electromagnetic radiation energy

International Nuclear Information System (INIS)

Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh

2014-01-01

The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding

Transparent Metal-Salt-Filled Polymeric Radiation Shields

Science.gov (United States)

Edwards, David; Lennhoff, John; Harris, George

2003-01-01

"COR-RA" (colorless atomic oxygen resistant -- radiation shield) is the name of a transparent polymeric material filled with x-ray-absorbing salts of lead, bismuth, cesium, and thorium. COR-RA is suitable for use in shielding personnel against bremsstrahlung radiation from electron-beam welding and industrial and medical x-ray equipment. In comparison with lead-foil and leaded-glass shields that give equivalent protection against x-rays (see table), COR-RA shields are mechanically more durable. COR-RA absorbs not only x-rays but also neutrons and rays without adverse effects on optical or mechanical performance. The formulation of COR-RA with the most favorable mechanical-durability and optical properties contains 22 weight percent of bismuth to absorb x-rays, plus 45 atomic percent hydrogen for shielding against neutrons.

Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

International Nuclear Information System (INIS)

Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

2015-01-01

As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

Shielding structure analysis for LSDS facility

Energy Technology Data Exchange (ETDEWEB)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

Shielding structure analysis for LSDS facility

International Nuclear Information System (INIS)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong

2014-01-01

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization

Shielding requirements for particle bed propulsion systems

Science.gov (United States)

Gruneisen, S. J.

1991-06-01

Nuclear Thermal Propulsion systems present unique challenges in reliability and safety. Due to the radiation incident upon all components of the propulsion system, shielding must be used to keep nuclear heating in the materials within limits; in addition, electronic control systems must be protected. This report analyzes the nuclear heating due to the radiation and the shielding required to meet the established criteria while also minimizing the shield mass. Heating rates were determined in a 2000 MWt Particle Bed Reactor (PBR) system for all materials in the interstage region, between the reactor vessel and the propellant tank, with special emphasis on meeting the silicon dose criteria. Using a Lithium Hydride/Tungsten shield, the optimum shield design was found to be: 50 cm LiH/2 cm W on the axial reflector in the reactor vessel and 50 cm LiH/2 cm W in a collar extension of the inside shield outside of the pressure vessel. Within these parameters, the radiation doses in all of the components in the interstage and lower tank regions would be within acceptable limits for mission requirements.

Shielding walls against ionizing radiation. Lead bricks

International Nuclear Information System (INIS)

1993-04-01

The standard contains specifications for the shape and requirements set for lead bricks such that they can be used to construct radiation-shielding walls according to the building kit system. The dimensions of the bricks are selected in such a way as to permit any modification of the length, height and thickness of said shielding walls in units of 50 mm. The narrow side of the lead bricks juxtaposed to one another in a wall construction to shield against radiation have to form prismatic grooves and tongues: in this way, direct penetration by radiation is prevented. Only cuboid bricks (serial nos. 55-60 according to Table 10) do not have prismatic tongues and grooves. (orig.) [de

Radiation shielding aspects for long manned mission to space - Criteria, survey study and preliminary model

International Nuclear Information System (INIS)

Sztejnberg, M.; Xiao, S.; Satvat, N.; Limon, F.; Hopkins, J.; Jevremovic, T.; T. Jevremovic)

2006-01-01

The prospect of manned space missions out side Earth's or bit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is there fore obvious that nuclear power production in space is a must. On the other hand, recent considerations to send a man to the Moon for a long stay would require a stable, secured, and safe source of energy (there is hardly anything beyond nuclear power that would provide a useful and reliably safe sustainable supply of energy). National Aeronautics and Space Administration (NASA) anticipates that the mass of a shielding material required for long travel to Mars is the next major design driver. In 2006 NASA identified a need to assess and evaluate potential gaps in existing knowledge and understanding of the level and types of radiation critical to astronauts' health during the long travel to Mars and to start a comprehensive study related to the shielding design of a spacecraft finding the conditions for the mitigation of radiation components contributing to the doses beyond accepted limits. In order to reduce the overall space craft mass, NASA is looking for the novel, multi-purpose and multi-functional materials that will provide effective shielding of the crew and electronics on board. The Laboratory for Neutronics and Geometry Computation in the School of Nuclear Engineering at Purdue University led by Prof. Tatjana Jevremovic began in 2004 the analytical evaluations of different lightweight materials. The preliminary results of the design survey study are presented in this paper. (author)

Radiation shielding aspects for long manned mission to space: Criteria, survey study, and preliminary model

Directory of Open Access Journals (Sweden)

Sztejnberg Manuel

2006-01-01

Full Text Available The prospect of manned space missions outside Earth's orbit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is therefore obvious that nuclear power production in space is a must. On the other hand, recent considerations to send a man to the Moon for a long stay would require a stable, secured and safe source of energy (there is hardly anything beyond nuclear power that would provide a useful and reliably safe sustainable supply of energy. National Aeronautics and Space Administration (NASA anticipates that the mass of a shielding material required for long travel to Mars is the next major design driver. In 2006 NASA identified a need to assess and evaluate potential gaps in existing knowledge and understanding of the level and types of radiation critical to astronauts' health during the long travel to Mars and to start a comprehensive study related to the shielding design of a spacecraft finding the conditions for the mitigation of radiation components contributing to the doses beyond accepted limits. In order to reduce the overall space craft mass, NASA is looking for the novel, multi-purpose and multi-functional materials that will provide effective shielding of the crew and electronics on board. The Laboratory for Neutronics and Geometry Computation in the School of Nuclear Engineering at Purdue University led by Prof. Tatjana Jevremović began in 2004 the analytical evaluations of different lightweight materials. The preliminary results of the design survey study are presented in this paper.

Guide to verification and validation of the SCALE-4 radiation shielding software

Energy Technology Data Exchange (ETDEWEB)

Broadhead, B.L.; Emmett, M.B.; Tang, J.S.

1996-12-01

Whenever a decision is made to newly install the SCALE radiation shielding software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the radiation shielding software in a version of SCALE-4. This report provides documentation of sample problems which are recommended for use in the V&V of the SCALE-4 system for all releases. The results reported in this document are from the SCALE-4.2P version which was run on an IBM RS/6000 work-station. These results verify that the SCALE-4 radiation shielding software has been correctly installed and is functioning properly. A set of problems for use by other shielding codes (e.g., MCNP, TWOTRAN, MORSE) performing similar V&V are discussed. A validation has been performed for XSDRNPM and MORSE-SGC6 utilizing SASI and SAS4 shielding sequences and the SCALE 27-18 group (27N-18COUPLE) cross-section library for typical nuclear reactor spent fuel sources and a variety of transport package geometries. The experimental models used for the validation were taken from two previous applications of the SASI and SAS4 methods.

Guide to verification and validation of the SCALE-4 radiation shielding software

International Nuclear Information System (INIS)

Broadhead, B.L.; Emmett, M.B.; Tang, J.S.

1996-12-01

Whenever a decision is made to newly install the SCALE radiation shielding software on a computer system, the user should run a set of verification and validation (V ampersand V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V ampersand V in that it specifies test cases to run and gives expected results. The report describes the V ampersand V that has been performed for the radiation shielding software in a version of SCALE-4. This report provides documentation of sample problems which are recommended for use in the V ampersand V of the SCALE-4 system for all releases. The results reported in this document are from the SCALE-4.2P version which was run on an IBM RS/6000 work-station. These results verify that the SCALE-4 radiation shielding software has been correctly installed and is functioning properly. A set of problems for use by other shielding codes (e.g., MCNP, TWOTRAN, MORSE) performing similar V ampersand V are discussed. A validation has been performed for XSDRNPM and MORSE-SGC6 utilizing SASI and SAS4 shielding sequences and the SCALE 27-18 group (27N-18COUPLE) cross-section library for typical nuclear reactor spent fuel sources and a variety of transport package geometries. The experimental models used for the validation were taken from two previous applications of the SASI and SAS4 methods

Shielding repair and comprehensive safety inspection of the nuclear-powered ship Mutsu

International Nuclear Information System (INIS)

Takada, Etsuo

1982-01-01

Eight years after the radiation leakage accident, the nuclear-powered ship Mutsu returned again to its home port Ominato. During the period, for four years, the n.s. Mutsu was subjected to shielding repair and comprehensive inspection at Sasebo port. In the future, the ship will start on experimental navigation after its functional and power-up tests. The works of shielding repair and the comprehensive inspection with subsequent repair are described in technical aspects. The basic policy of the repair was two points, i.e. the usage of shielding materials excellent in shielding capacity, less in radioactivation and enduring operating temperature, and structural strength resisting ship-hull acceleration, shock and vibration, enabling easy maintenance and inspection. Comprehensive inspection was made on not only machinery integrity but also the design itself. (Mori, K.)

Discussions for the shielding materials of synchrotron radiation beamline hutches

International Nuclear Information System (INIS)

Asano, Y.

2006-01-01

Many synchrotron radiation facilities are now under operation such as E.S.R.F., APS, and S.P.ring-8. New facilities with intermediated stored electron energy are also under construction and designing such as D.I.A.M.O.N.D., S.O.L.E.I.L., and S.S.R.F.. At these third generation synchrotron radiation facilities, the beamline shielding as well as the bulk shield is very important for designing radiation safety because of intense and high energy synchrotron radiation beam. Some reasons employ lead shield wall for the synchrotron radiation beamlines. One is narrow space for the construction of many beamlines at the experimental hall, and the other is the necessary of many movable mechanisms at the beamlines, for examples. Some cases are required to shield high energy neutrons due to stored electron beam loss and photoneutrons due to gas Bremsstrahlung. Ordinary concrete and heavy concrete are coming up to shield material of synchrotron radiation beamline hutches. However, few discussions have been performed so far for the shielding materials of the hutches. In this presentation, therefore, we will discuss the characteristics of the shielding conditions including build up effect for the beamline hutches by using the ordinary concrete, heavy concrete, and lead for shielding materials with 3 GeV and 8 GeV class synchrotron radiation source. (author)

Development of epoxy resin-type neutron shielding materials (I)

Energy Technology Data Exchange (ETDEWEB)

Cho, Soo Haeng; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy

1997-12-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear /radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various materials properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. (author). 31 refs., 22 tabs., 17 figs.

Foam-Reinforced Polymer Matrix Composite Radiation Shields, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — New and innovative lightweight radiation shielding materials are needed to protect humans in future manned exploration vehicles. Radiation shielding materials are...

Mechanical and radiation shielding properties of mortars with additive fine aggregate mine waste

International Nuclear Information System (INIS)

Gallala, Wissem; Hayouni, Yousra; Gaied, Mohamed Essghaier; Fusco, Michael; Alsaied, Jasmin; Bailey, Kathryn; Bourham, Mohamed

2017-01-01

Highlights: • Effectiveness of mine waste as additive fine aggregate has been investigated. • Experimental results are verified by computationally from composition of synthesized samples. • Work focuses on shielding materials for nuclear systems including spent fuel storage and drycasks. - Abstract: Incorporation of barite-fluorspar mine waste (BFMW) as a fine aggregate additive has been investigated for its effect on the mechanical and shielding properties of cement mortar. Several mortar mixtures were prepared with different proportions of BFMW ranging from 0% to 30% as fine aggregate replacement. Cement mortar mixtures were evaluated for density, compressive and tensile strengths, and gamma ray radiation shielding. The results revealed that the mortar mixes containing 25% BFMW reaches the highest compressive strength values, which exceeded 50 MPa. Evaluation of gamma-ray attenuation was both measured by experimental tests and computationally calculated using MicroShield software package, and results have shown that using BFMW aggregates increases attenuation coefficient by about 20%. These findings have demonstrated that the mine waste can be suitably used as partial replacement aggregate to improve radiation shielding as well as to reduce the mortar and concrete costs.

γ-ray shielding behaviors of some nuclear engineering materials

International Nuclear Information System (INIS)

Mann, Kulwinder Singh

2017-01-01

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well

γ-ray shielding behaviors of some nuclear engineering materials

Energy Technology Data Exchange (ETDEWEB)

Mann, Kulwinder Singh [Dept. of Physics, D.A.V. College, Punjab (India)

2017-06-15

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

γ-Ray Shielding Behaviors of Some Nuclear Engineering Materials

Directory of Open Access Journals (Sweden)

Kulwinder Singh Mann

2017-06-01

Full Text Available The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM. The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB of six glass samples (transparent NEM were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV and optical thickness (OT. The study was performed by computing various γ-ray shielding parameters (GSP such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

A history of radiation shielding of x-ray therapy rooms

International Nuclear Information System (INIS)

McGinley, P.H.; Miner, M.S.

1996-01-01

In this report the history of shielding for radiation treatment rooms is traced from the time of the discovery of x rays to the present. During the early part of the twentieth century the hazards from ionizing radiation were recognized and the use of lead and other materials became common place for shielding against x rays. Techniques for the calculation of the shield thickness needed for x ray protection were developed in the 1920's, and shielding materials were characterized in terms of the half value layer or simple exponential factors. At the same time, better knowledge of the interaction between radiation and matter was acquired. With the development of high energy medical accelerators after 1940, new and more complex shielding problems had to be addressed. Recently, shielding requirements have become more stringent as standards for exposure of personnel and the general public have been reduced. The art of shielding of radiation treatment facilities is still being developed, and the need for a revision of the reports on shielding of medical accelerators from the National Council on Radiation Protection and Measurements is emphasized in this article. (author). 61 Refs., 3 Tabs

REPOSITORY RADIATION SHIELDING DESIGN GUIDE

International Nuclear Information System (INIS)

M. Haas; E.M. Fortsch

1997-01-01

The scope of this document includes radiation safety considerations used in the design of facilities for the Yucca Mountain Site Characterization Project (YMP). The purpose of the Repository Radiation Shielding Design Guide is to document the approach used in the radiological design of the Mined Geologic Disposal System (MGDS) surface and subsurface facilities for the protection of workers, the public, and the environment. This document is intended to ensure that a common methodology is used by all groups that may be involved with Radiological Design. This document will also assist in ensuring the long term survivability of the information basis used for radiological safety design and will assist in satisfying the documentation requirements of the licensing body, the Nuclear Regulatory Commission (NRC). This design guide provides referenceable information that is current and maintained under the YMP Quality Assurance (QA) Program. Furthermore, this approach is consistent with maintaining continuity in spite of a changing design environment. This approach also serves to ensure common inter-disciplinary interpretation and application of data

REPOSITORY RADIATION SHIELDING DESIGN GUIDE

Energy Technology Data Exchange (ETDEWEB)

M. Haas; E.M. Fortsch

1997-09-12

The scope of this document includes radiation safety considerations used in the design of facilities for the Yucca Mountain Site Characterization Project (YMP). The purpose of the Repository Radiation Shielding Design Guide is to document the approach used in the radiological design of the Mined Geologic Disposal System (MGDS) surface and subsurface facilities for the protection of workers, the public, and the environment. This document is intended to ensure that a common methodology is used by all groups that may be involved with Radiological Design. This document will also assist in ensuring the long term survivability of the information basis used for radiological safety design and will assist in satisfying the documentation requirements of the licensing body, the Nuclear Regulatory Commission (NRC). This design guide provides referenceable information that is current and maintained under the YMP Quality Assurance (QA) Program. Furthermore, this approach is consistent with maintaining continuity in spite of a changing design environment. This approach also serves to ensure common inter-disciplinary interpretation and application of data.

Radiation shielding in dental radiography

International Nuclear Information System (INIS)

Stenstroem, B.; Rehnmark-Larsson, S.; Julin, P.; Richter, S.

1983-01-01

The protective effect in the thyroid region from different types of radiation shieldings at intraoral radiography has been studied as well as the reduction of the absorbed dose to the sternal and the gonadal regions. The shieldings tested were five different types of leaded aprons, of which three had an attached leaded collar and the other two were used in combination with separate soft leaded collars. Furthermore one of the soft leaded collars and an unflexible horizontal leaded shield were tested separately. Two dental x-ray machines of 60 and 65 kVp with rectangular and circular tube collimators were used. The exposure time corresponded to speed group E film. The absorbed doses were measured with two ionization chambers. No significant difference in the protective effect in the thyroid gland could be found between the different types of radiation shieldings. There was a dose reduction by approximately a factor of 2 to the thyroid region down to 0.08 mGy per full survey using parallelling technique, and below 0.001 mGy per single bitewing exposure. The shieldings reduced the thyroid dose using bisecting-angle technique by a factor of 5 down to 0.15 mGy per full survey (20 exposures). In the sternal region the combinations of apron and collar reduced the absorbed dose from a full survey to below 2 μGy compared with 18 μGy (parallelling) and 31 μGy (biscting-angle) without any shielding. With the horizontal leaded shield a reduction by a factor of 6 was obtained but no significant sternal dose reduction could be detected from the soft collar alone. The gonadal dose could be reduced by a factor of 10 with the horizontal leaded shield, parallelling technique and circular collimator. Using leaded aprons the gonadal dose was approximately one per cent of the dose without any shielding, i.e. below 0.01 μGy per single intraoral exposure. (Authors)

Radiation protection and shielding design - Strengthening the link

International Nuclear Information System (INIS)

Hobson, J.; Cooper, A.

2005-01-01

The improvement in quality and flexibility of shielding methods and data has been progressive and beneficial in opening up new opportunities for optimising radiation protection in design. The paper describes how these opportunities can best be seized by taking a holistic view of radiation protection, with shielding design being an important component part. This view is best achieved by enhancing the role of 'shielding assessors' so that they truly become 'radiation protection designers'. The increase in speed and efficiency of shielding calculations has been enormous over the past decades. This has raised the issue of how the assessor's time now can be best utilised; pursuing ever greater precision and accuracy in shielding/dose assessments, or improving the contribution that shielding assessment makes to radiological protection and cost-effective design. It is argued in this paper that the latter option is of great importance and will give considerable benefits. Shielding design needs to form part of a larger radiation protection perspective based on a deep understanding/appreciation of the opportunities and constraints of operators and designers, enabling minimal design iterations, cost optimisation of alternative designs (with a 'lifetime' perspective) and improved realisation of design intent in operations. The future of shielding design development is argued to be not in improving the 'tool-kit', but in enhanced understanding of the 'product' and the 'process' for achieving it. The holistic processes being developed in BNFL to realise these benefits are described in the paper and will be illustrated by case studies. (authors)

Preliminary radiation shielding design for BOOMERANG

International Nuclear Information System (INIS)

Donahue, Richard J.

2002-01-01

Preliminary radiation shielding specifications are presented here for the 3 GeV BOOMERANG Australian synchrotron light source project. At this time the bulk shield walls for the storage ring and injection system (100 MeV Linac and 3 GeV Booster) are considered for siting purposes

Radiation shielding issues on the FMIT

International Nuclear Information System (INIS)

Burke, R.J.; Davis, A.A.; Huang, S.; Morford, R.J.

1981-05-01

The Fusion Materials Irradiation Test Facility (FMIT) is being built to study neutron radiation effects in candidate fusion reactor materials. The FMIT will yield high fluence data in a fusion-like neutron radiation environment produced by the interaction of a 0.1A, 35 MeV deuteron beam with a flowing lithium target. The design of the facility as a whole is driven by a high availability requirement. The variety of radiation environments in the facility requires the use of diverse and extensive shielding. Shielding design throughout the FMIT must accommodate the need for maintenance and operations access while providing adequate personnel and equipment protection

PMMA/MWCNT nanocomposite for proton radiation shielding applications

Science.gov (United States)

Li, Zhenhao; Chen, Siyuan; Nambiar, Shruti; Sun, Yonghai; Zhang, Mingyu; Zheng, Wanping; Yeow, John T. W.

2016-06-01

Radiation shielding in space missions is critical in order to protect astronauts, spacecraft and payloads from radiation damage. Low atomic-number materials are efficient in shielding particle-radiation, but they have relatively weak material properties compared to alloys that are widely used in space applications as structural materials. However, the issues related to weight and the secondary radiation generation make alloys not suitable for space radiation shielding. Polymers, on the other hand, can be filled with different filler materials for reinforcement of material properties, while at the same time provide sufficient radiation shielding function with lower weight and less secondary radiation generation. In this study, poly(methyl-methacrylate)/multi-walled carbon nanotube (PMMA/MWCNT) nanocomposite was fabricated. The role of MWCNTs embedded in PMMA matrix, in terms of radiation shielding effectiveness, was experimentally evaluated by comparing the proton transmission properties and secondary neutron generation of the PMMA/MWCNT nanocomposite with pure PMMA and aluminum. The results showed that the addition of MWCNTs in PMMA matrix can further reduce the secondary neutron generation of the pure polymer, while no obvious change was found in the proton transmission property. On the other hand, both the pure PMMA and the nanocomposite were 18%-19% lighter in weight than aluminum for stopping the protons with the same energy and generated up to 5% fewer secondary neutrons. Furthermore, the use of MWCNTs showed enhanced thermal stability over the pure polymer, and thus the overall reinforcement effects make MWCNT an effective filler material for applications in the space industry.

Problems of the power plant shield optimization

International Nuclear Information System (INIS)

Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

1981-01-01

General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

Radiation dose reduction by water shield

International Nuclear Information System (INIS)

Zeb, J.; Arshed, W.; Ahmad, S.S.

2007-06-01

This report is an operational manual of shielding software W-Shielder, developed at Health Physics Division (HPD), Pakistan Institute of Nuclear Science and Technology (PINSTECH), Pakistan Atomic Energy Commission. The software estimates shielding thickness for photons having their energy in the range 0.5 to 10 MeV. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. Water is used as a shielding material in this software. The software is helpful in estimating the water thickness for safe handling, storage of gamma emitting radionuclide. (author)

Shielding modification and safety review on the nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Osanai, Masao

1978-01-01

The Japan Atomic Energy Commission (JAEC) called on the Japan Nuclear Ship Development Agency (JNSDA) for shielding modification and safety review on the nuclear ship ''Mutsu'', and JNSDA has conducted the research and development (R and D) to meet the request of JAEC for the above two items. Concerning the shield modification, the following matters are described: the study on the cause of radiation leakage which was concluded to the fast neutron streaming, the conceptual design for this modification, the mock up experiment for shielding utilizing JRR-4, the basic design following on the conceptual design, including the detailed drawings of the modified construction and the shielding analysis using RADHEAT-V3 code, and the relating experiments such as the heat transfer test of the primary shielding structure and the test of strength in stranding. As for the safety review, the survey of the troubles and the technical problems having been experienced in the light water reactor plants of land use, for example, fuel integrity, stress corrosion cracking and the leakage of steam generator tubes, the revision of the design so as to adapt to current safety standards and regulations, for example, in-service inspection, the setting of additional leak detectors in the primary cooling system, the modification of emergeney filters, etc., and the review of the design and construction corresponding to recent R and D works, such as re-evaluation of the core design, cooling capability of natural circulation, thermal stress analysis of main pipings, and the evaluation of ECCS performance are presented . (Nakai, Y.)

High ionization radiation field remote visualization device - shielding requirements

International Nuclear Information System (INIS)

Fernandez, Antonio P. Rodrigues; Omi, Nelson M.; Silveira, Carlos Gaia da; Calvo, Wilson A. Pajero

2011-01-01

The high activity sources manipulation hot-cells use special and very thick leaded glass windows. This window provides a single sight of what is being manipulated inside the hot-cell. The use of surveillance cameras would replace the leaded glass window, provide other sights and show more details of the manipulated pieces, using the zoom capacity. Online distant manipulation may be implemented, too. The limitation is their low ionizing radiation resistance. This low resistance also limited the useful time of robots made to explore or even fix problematic nuclear reactor core, industrial gamma irradiators and high radioactive leaks. This work is a part of the development of a high gamma field remote visualization device using commercial surveillance cameras. These cameras are cheap enough to be discarded after the use for some hours of use in an emergency application, some days or some months in routine applications. A radiation shield can be used but it cannot block the camera sight which is the shield weakness. Estimates of the camera and its electronics resistance may be made knowing each component behavior. This knowledge is also used to determine the optical sensor type and the lens material, too. A better approach will be obtained with the commercial cameras working inside a high gamma field, like the one inside of the IPEN Multipurpose Irradiator. The goal of this work is to establish the radiation shielding needed to extend the camera's useful time to hours, days or months, depending on the application needs. (author)

PEP radiation shielding tests in SLAC A Beam

International Nuclear Information System (INIS)

Ash, W.; DeStaebler, H.; Harris, J.; Jenkins, T.; Murray, J.

1977-09-01

Radiation shielding tests designed to simulate possible conditions in and around the PEP experimental halls were conducted. The SLAC A Beam was targeted in the block tunnel at a point about midway between End Station A and Beam Dump East. At that site it was relatively easy to rearrange the concrete block structure to simulate the various shielding configurations under consideration for PEP. Extensive surveys of neutron and ionizing radiation were made. Complete results of the shielding tests are given

Effect of background radiation shielding on natural radioactivity distribution measurement with imaging plate

International Nuclear Information System (INIS)

Mori, C.; Suzuki, T.; Koido, S.; Uritani, A.; Miyahara, H.; Yanagida, K.; Miyahara, J.; Takahashi, K.

1996-01-01

Distribution images of natural radioactivity contained in various natural materials such as vegetable, animal meat and pottery work can be obtained with an imaging plate which has high sensitivity for nuclear radiations. For such very low levels of radioactivity, natural background radiations must be reduced using a shielding box. The lining, on the inside of the box, with low atomic number material such as acrylic resin is very effective in reducing electrons, β-rays and low energy X- and γ-rays emitted from the inner surface of the shielding material. Some images of natural radioactivity distribution were obtained and the radioactivity, mainly 40 K, contained in natural materials was measured by using an HPGe detector and also the imaging plate itself. (orig.)

Electrically nonconductive shield for electric equipment generating ionizing radiation

International Nuclear Information System (INIS)

Aitken, D.

1979-01-01

As a radiation protection shield there is proposed a nonconductive shield fabricated from epoxides or other plastics material and containing finely dispersed radiation absorbing metal. It is to be designed in such a way that it lies in the range of a high electric gradient in the equipment, close to the radiation-producing component. As suitable metals there are mentioned tin, tungsten, and lead resp. their oxides. As an example there is used an X-ray shielding. (RW) 891 RW/RW 892 MKO [de

Radiation-shielding transparent material

International Nuclear Information System (INIS)

Kusumeki, Asao.

1983-01-01

Purpose : To obtain radiation-shielding transparent material having a high resistivity to the radioactive rays or light irradiation which is greater at least by two digits as compared with lead glass. Constitution : The shielding material is composed of a saturated aqueous solution zinc iodide. Zinc iodide (specific gravity of 4.2) is dissolved by 430 g into 100 cc of water at a temperature of 20 0 C and forms a heavy liquid with a specific gravity of 2.80. The radiation length of the heavy liquid is 3.8 cm which is 1.5 times as large as lead glass. The light transmission is greater than 95% in average. Furthermore, by adding hypophosphorous acid as a reducing agent to the aqueous solution of the lead iodide, the material is stabilized against the irradiation of light or radioactive rays and causes no discoloration for a long time. (Moriyama, K.)

Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

Energy Technology Data Exchange (ETDEWEB)

Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

2017-03-15

Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

Measurement of TFTR D-T radiation shielding efficiency

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione G.; Elwood, S.

1994-01-01

High power D-T fusion reactor designs presently exhibit complex geometric and material density configurations. Simulations of the radiation shielding required for safe operation and full compliance with all regulatory requirements must include sufficient margin to accommodate uncertainties in material properties and distributions, uncertainties in the final configurations, and uncertainties in approximations employing the homogenization of complex geometries. Measurements of radiation shielding efficiency performed in a realistic D-T tokamak environment can provide empirical guidance for simulating safe, efficient, and cost effective shielding systems for future high power fusion reactors. In this work, the authors present the results of initial measurements of the TFTR radiation shielding efficiency during high power D-T operations with record neutron yields. The TFTR design objective is to limit the total dose-equivalent at the nearest PPPL property lines from all radiation pathways to 10 mrem per calendar year. Compliance with this design objective over a calendar year requires measurements in the presence of typical site backgrounds of about 80 mrem per year

Active Radiation Shield, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — DEC-Shield technology offers the means to generate electric power from cosmic radiation sources and fuse dissimilar systems and functionality into a structural...

Study and application of the ANISN and DOT 3.5 codes to problems in nuclear radiation shielding

International Nuclear Information System (INIS)

Otto, A.C.

1983-01-01

The application of the Sn transport codes ANISN and DOT 3.5 to problems in radiation shielding is reviewed. In addition, a large array of codes involved in radiation shielding calculations is described and applied in this work. The ANISN and DOT 3.5 codes solve the multigroup transport equation in plane, cylindrical and spherical geometries, the first in one dimension and the second in two dimensions, by using the Sn approximation and were designed to solve coupled neutron-photon transport problems commonly found in reactor shielding calculations. In this work the numerical methods used in these codes are reviewed and their basic application to deep-penetration and void problems is discussed. Benchmark problems are solved by employing the array of codes previously mentioned. In particular, the ability of the ISOFLUXO program coupled to the DOT 3.5 code of mapping contours of regions with approximately the same scalar fluxes is illustrated, showing that they can be efficiently used in shielding analysis. (Author) [pt

Passive radiation shielding considerations for the proposed space elevator

Science.gov (United States)

Jorgensen, A. M.; Patamia, S. E.; Gassend, B.

2007-02-01

The Earth's natural van Allen radiation belts present a serious hazard to space travel in general, and to travel on the space elevator in particular. The average radiation level is sufficiently high that it can cause radiation sickness, and perhaps death, for humans spending more than a brief period of time in the belts without shielding. The exact dose and the level of the related hazard depends on the type or radiation, the intensity of the radiation, the length of exposure, and on any shielding introduced. For the space elevator the radiation concern is particularly critical since it passes through the most intense regions of the radiation belts. The only humans who have ever traveled through the radiation belts have been the Apollo astronauts. They received radiation doses up to approximately 1 rem over a time interval less than an hour. A vehicle climbing the space elevator travels approximately 200 times slower than the moon rockets did, which would result in an extremely high dose up to approximately 200 rem under similar conditions, in a timespan of a few days. Technological systems on the space elevator, which spend prolonged periods of time in the radiation belts, may also be affected by the high radiation levels. In this paper we will give an overview of the radiation belts in terms relevant to space elevator studies. We will then compute the expected radiation doses, and evaluate the required level of shielding. We concentrate on passive shielding using aluminum, but also look briefly at active shielding using magnetic fields. We also look at the effect of moving the space elevator anchor point and increasing the speed of the climber. Each of these mitigation mechanisms will result in a performance decrease, cost increase, and technical complications for the space elevator.

Radiation shielding analysis

International Nuclear Information System (INIS)

Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.

1982-01-01

The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)

Development of Computer Program for Analysis of Irregular Non Homogenous Radiation Shielding

International Nuclear Information System (INIS)

Bang Rozali; Nina Kusumah; Hendro Tjahjono; Darlis

2003-01-01

A computer program for radiation shielding analysis has been developed to obtain radiation attenuation calculation in non-homogenous radiation shielding and irregular geometry. By determining radiation source strength, geometrical shape of radiation source, location, dimension and geometrical shape of radiation shielding, radiation level of a point at certain position from radiation source can be calculated. By using a computer program, calculation result of radiation distribution analysis can be obtained for some analytical points simultaneously. (author)

Organ dose assessment of nuclear medicine practitioners using L-block shielding device for handing diagnostic radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Kang, Se Sik; Kim, Jung Hoon [Dep. of Radiological Science, College of Health Science, Catholic University of Pusan, Busan (Korea, Republic of); Cho, Yong In [Dept. of Diagnostic Radiology, Dongnam Institute of Radiological and Medical Science, Busan (Korea, Republic of)

2017-03-15

In the case of nuclear medicine practitioners in medical institutions, a wide range of exposure dose to individual workers can be found, depending on the type of source, the amount of radioactivity, and the use of shielding devices in handling radioactive isotopes. In this regard, this study evaluated the organ dose on practitioners as well as the dose reduction effect of the L-block shielding device in handling the diagnostic radiation source through the simulation based on the Monte Carlo method. As a result, the distribution of organ dose was found to be higher as the position of the radiation source was closer to the handling position of a practitioner, and the effective dose distribution was different according to the ICRP tissue weight. Furthermore, the dose reduction effect according to the L-block thickness tended to decrease, which showed the exponential distribution, as the shielding thickness increased. The dose reduction effect according to each radiation source showed a low shielding effect in proportion to the emitted gamma ray energy level.

Measurement of radiation shielding properties of polymer composites by using HPGe detector

International Nuclear Information System (INIS)

Gupta, Anil; Pillay, H.C.M.; Kale, P.K.; Datta, D.; Suman, S.K.; Gover, V.

2014-01-01

Lead is the most common radiation shield and its composite with polymers can be used as flexible radiation shields for different applications. However, lead is very hazardous and has been found to be associated with neurological disorders, kidney failure and hematotoxicity. Lead free radiation shield material has been developed by synthesizing radiation cross linked PDMS/Bi 2 O 3 polymer composites. In order to have a lead free radiation shield the relevant shielding properties such as linear attenuation, half value thickness (HVT) and tenth value thickness (TVT) have been measured by using HPGe detector. The present study describes the methodology of measurement of the shielding properties of the lead free shield material. In the measurement gamma energies such as 59.537 keV ( 241 Am), 122.061 keV and 136.474 keV ( 57 Co) are taken into consideration

Radiation protection in nuclear medicine

International Nuclear Information System (INIS)

Seeburrun, V.

2013-04-01

Radiation protection in nuclear medicine in this project is concerned with the reduction of doses to workers, patients and members of the public. Protection of workers is achieved by adopting good personal habits, good housekeeping, proper use of personal protective devices and equipment, attend training and have continuous education. Exposure to radiation of workers and the members of the public are minimised by proper management of radioactive waste and safe transport of radioactive material. The design and shielding of a nuclear medicine department shall further provide for the protection of the worker, the patient and the general public. Protection of patient is achieved by justifying the procedure, delivering the minimum radiation dose possible to the patient while obtaining the best image quality and applying guidance levels. Special considerations shall be given to pregnant and breast-feeding patients. Quality assurance programme through image quality, radiopharmaceutical quality and patient records on nuclear medicine procedures shall provide assurance to the patient. (au)

ANS shielding standards for light-water reactors

International Nuclear Information System (INIS)

Trubey, D.K.

1982-01-01

The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

International Nuclear Information System (INIS)

Dunley, Leonardo Souza

2002-01-01

The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

Practical radiation shielding for biomedical research

International Nuclear Information System (INIS)

Klein, R.C.; Reginatto, M.; Party, E.; Gershey, E.L.

1990-01-01

This paper reports on calculations which exist for estimating shielding required for radioactivity; however, they are often not applicable for the radionuclides and activities common in biomedical research. A variety of commercially available Lucite shields are being marketed to the biomedical community. Their advertisements may lead laboratory workers to expect better radiation protection than these shields can provide or to assume erroneously that very weak beta emitters require extensive shielding. The authors have conducted a series of shielding experiments designed to simulate exposures from the amounts of 32 P, 51 Cr and 125 I typically used in biomedical laboratories. For most routine work, ≥0.64 cm of Lucite covered with various thicknesses of lead will reduce whole-body occupational exposure rates of < 1mR/hr at the point of contact

Estimation of temperature distribution in a reactor shield

International Nuclear Information System (INIS)

Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

1989-01-01

Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

Evaluation of usability of the shielding effect for thyroid shield for peripheral dose during whole brain radiation therapy

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Sic; Park, Ju Kyeong; Lee, Seung Hun; Kim, Yang Su; Lee, Sun Young; Cha, Seok Yong [Dept. of Radiation Oncology, Chonbuk National University Hospital, Jeonju (Korea, Republic of)

2014-12-15

To reduce the radiation dose to the thyroid that is affected to scattered radiation, the shield was used. And we evaluated the shielding effect for the thyroid during whole brain radiation therapy. To measure the dose of the thyroid, 300cGy were delivered to the phantom using a linear accelerator(Clinac iX VARIAN, USA.)in the way of the 6MV X-ray in bilateral. To measure the entrance surface dose of the thyroid, five glass dosimeters were placed in the 10th slice's surface of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. In the same location, to measure the depth dose of the thyroid, five glass dosimeters were placed in the 10th slice by 2.5 cm depth of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. Entrance surface dose of the thyroid were respectively 44.89 mGy at the unshield, 36.03 mGy at the bismuth shield, 31.03 mGy at the 0.5 mmPb shield and 23.21 mGy at a self-made 1.0 mmPb shield. In addition, the depth dose of the thyroid were respectively 36.10 mGy at the unshield, 34.52 mGy at the bismuth shield, 32.28 mGy at the 0.5 mmPb shield and 25.50 mGy at a self-made 1.0 mmPb shield. The thyroid was affected by the secondary scattering dose and leakage dose outside of the radiation field during whole brain radiation therapy. When using a shield in the thyroid, the depth dose of thyroid showed 11-30% reduction effect and the surface dose of thyroid showed 20-48% reduction effect. Therefore, by using the thyroid shield, it is considered to effectively protect the thyroid and can perform the treatment.

Shielding of medically used proton accelerators; Abschirmung von medizinisch genutzten Protonenbeschleunigern

Energy Technology Data Exchange (ETDEWEB)

Ewen, Klaus

2014-10-01

In several standards of the standards committee radiology (NRA) the shielding of proton accelerators (cyclotrons) for medical utilization is described. Proton beams can be used in nuclear medicine for PET (proton emission tomography) isotope production or for radiotherapeutic use. The dominating radiation from proton induced nuclear reactions is fast neutron radiation. The calculation procedure for appropriate shielding measures according to the NAR standards is described step-by-step. AN adequate shielding of fast neutrons is also sufficient for the generated gamma radiation.

Combination thermal and radiation shield for well logging apparatus

International Nuclear Information System (INIS)

Wilson, B.F.

1984-01-01

A device for providing both thermal protection and radiation shielding for components such as radiation detectors within a well logging instrument comprises a thermally insulative flask containing a weldment filled with a mass of eutectic material which undergoes a change of state e.g. melting at a temperature which will provide an acceptable thermal environment for such components for extended time periods. The eutectic material which is preferably a bismuth (58%)/tin (42%) alloy has a specific gravity (> 8.5) facilitating its use as a radiation shield and is distributed around the radiation detectors so as to selectively impede the impinging of the detectors by radiation. The device is incorporated in a skid of a well logging instrument for measuring γ backscatter. A γ source is located either above or within the protective shielding. (author)

Adaptation of radiation shielding code to space environment

International Nuclear Information System (INIS)

Okuno, Koichi; Hara, Akihisa

1992-01-01

Recently, the trend to the development of space has heightened. To the development of space, many problems are related, and as one of them, there is the protection from cosmic ray. The cosmic ray is the radiation having ultrahigh energy, and there was not the radiation shielding design code that copes with cosmic ray so far. Therefore, the high energy radiation shielding design code for accelerators was improved so as to cope with the peculiarity that cosmic ray possesses. Moreover, the calculation of the radiation dose equivalent rate in the moon base to which the countermeasures against cosmic ray were taken was simulated by using the improved code. As the important countermeasures for the safety protection from radiation, the covering with regolith is carried out, and the effect of regolith was confirmed by using the improved code. Galactic cosmic ray, solar flare particles, radiation belt, the adaptation of the radiation shielding code HERMES to space environment, the improvement of the three-dimensional hadron cascade code HETCKFA-2 and the electromagnetic cascade code EGS 4-KFA, and the cosmic ray simulation are reported. (K.I.)

Radiation-related impacts for nuclear plant physical modifications

International Nuclear Information System (INIS)

Sciacca, F.; Knudson, R.; Simion, G.; Baca, G.; Behling, H.; Behling, K.; Britz, W.; Cohen, S.

1989-10-01

The radiation fields in nuclear power plants present significant obstacles to accomplishing repairs and modifications to many systems and components in these plants. The NRC's generic cost estimating methodology attempts to account for radiation-related impacts by assigning values to the radiation labor productivity factor. This radiation labor productivity factor is then used as a multiplier on the greenfield or new nuclear plant construction labor to adjust for the actual operating plant conditions. The value assigned to the productivity factor is based on the work-site radiation levels. The relationship among ALARA practices, work-place radiation levels, and radiation-related cost impacts previously had not been adequately characterized or verified. The assumptions made concerning the use and application of radiation-reduction measures such as system decontamination and/or the use of temporary shielding can significantly impact estimates of both labor requirements and radiation exposure associated with a particular activity. Overall guidance was needed for analysts as to typical ALARA practices at nuclear power plants and the effects of these practices in reducing work-site dose rates and overall labor requirements. This effort was undertaken to better characterize the physical modification cost and radiological exposure impacts related to the radiation environment of the work place. More specifically, this work sought to define and clarify the quantitative relationships between or among: radiation levels and ALARA practices, such as the use of temporary shielding, decontamination efforts, or the use of robots and remote tools; radiation levels and labor productivity factors; radiation levels, in-field labor hours, and worker radiation exposure; radiation levels and health physics services costs; and radiation levels, labor hours, and anti-contamination clothing and equipment. 48 refs., 4 figs., 4 tabs

A study on radiation shield design of storage facility for low and intermediate level radioactive waste in Bangladesh

International Nuclear Information System (INIS)

Khan, JJahirul Haque

2005-02-01

Bangladesh has no nuclear power reactor but has only one 3 MW TRIGA Mark-II Research Reactor. The Bangladesh Atomic Energy commission (BAEC) operates a 3 MW TRIGA Mark-II Research Reactor and maintains not only the nuclear facilities at its Atomic Energy Research Establishment (AERE) at Savar (near Dhaka) but also the related radiation facilities the whole country. The main sources of radioactive wastes result from the use of sealed and unsealed radiation sources in medicine industry, research, agriculture, etc as well as from operation and maintenance of the nuclear facilities the whole country. As a result radioactive wastes are increasing day by day and these wastes are classified as low and intermediate level radioactive waste (LILW) following the radiation safety philosophy of IAEA recommendations in Bangladesh. Radioactive waste is very sensitive issue to public and environment from the hazardous standpoint of ionizing radiation. Therefore, storage facility of LILW is very essential for safe radioactive waste management in Bangladesh and in parallel: this study is of a great importance due to new installation of this storage facility in future. The basic objective of this study is to recommend the radiation shield design parameters of the installation of storage facility for low and intermediate level radioactive waste from the points of view of radiation safety and sensitivity analysis. The shield design of this installation has been carried out with the Monte Carlo Code MCNP4C and the point Kernel Code Micro Shield 5.05 respectively considering the ICRP-60 (1990) recommendations for occupational exposure limit (10 μ Sv/hr). For more safety purpose every equivalent dose rate at different positions of this installation is considered below 9 μ Sv/hr in this study. The radiation shield design parameters are recommended based on MCNP4C calculated results than those of Micro Shield due to more credible results and these parameters are: (I) 51 cm thickness of

PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

International Nuclear Information System (INIS)

Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

1990-01-01

1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

International Nuclear Information System (INIS)

Lee, Yoon Hee

2006-02-01

Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Hee

2006-02-15

Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

Shield structure for a nuclear reactor

International Nuclear Information System (INIS)

Rouse, C.A.; Simnad, M.T.

1979-01-01

An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

The Radiation Safety Information Computational Center (RSICC): A Resource for Nuclear Science Applications

Energy Technology Data Exchange (ETDEWEB)

Kirk, Bernadette Lugue [ORNL

2009-01-01

The Radiation Safety Information Computational Center (RSICC) has been in existence since 1963. RSICC collects, organizes, evaluates and disseminates technical information (software and nuclear data) involving the transport of neutral and charged particle radiation, and shielding and protection from the radiation associated with: nuclear weapons and materials, fission and fusion reactors, outer space, accelerators, medical facilities, and nuclear waste management. RSICC serves over 12,000 scientists and engineers from about 100 countries.

Radiation shielding quality assurance

Science.gov (United States)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

International Nuclear Information System (INIS)

Kirchner, J.

1984-01-01

Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

Radiation shielding for the Super Collider West Utility region

International Nuclear Information System (INIS)

Meinke, R.; Mokhov, N.; Orth, D.; Parker, B.; Plant, D.

1994-02-01

Shielding considerations in the 20 x 20-TeV Superconducting Super Collider are strongly correlated with detailed machine specifics in the various accelerator sections. The West Utility, the most complex area of the Collider, concentrates all the major accelerator subsystems in a single area. The beam loss rate and associated radiation levels in this region are anticipated to be quite high, and massive radiation shielding is therefore required to protect personnel, Collider components, and the environment. The challenging task of simultaneously optimizing accelerator design and radiation shielding, both of which are strongly influenced by subsystem design details, requires the integration of several complex simulation codes. To this end we have performed exhaustive hadronic shower simulations with the MARS12 program; detailed accelerator lattice and optics optimization via the SYNCH, MAD, and MAGIC codes; and extensive 3-D configuration modeling of the accelerator tunnel and subsystems geometries. Our technique and the non-trivial results from such a combined approach are presented here. An integrated procedure is found invaluable in developing cost-effective radiation shielding solutions

Nuclear reactor shield including magnesium oxide

International Nuclear Information System (INIS)

Rouse, C.A.; Simnad, M.T.

1981-01-01

An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

International Nuclear Information System (INIS)

Ueki, K.; Nariyama, N.; Ohashi, A.

2000-01-01

The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

International Nuclear Information System (INIS)

Meuschke, R.E.; Boyd, C.H.

1989-01-01

This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

Multi-objective optimization design method of radiation shielding

International Nuclear Information System (INIS)

Yang Shouhai; Wang Weijin; Lu Daogang; Chen Yixue

2012-01-01

Due to the shielding design goals of diversification and uncertain process of many factors, it is necessary to develop an optimization design method of intelligent shielding by which the shielding scheme selection will be achieved automatically and the uncertainties of human impact will be reduced. For economical feasibility to achieve a radiation shielding design for automation, the multi-objective genetic algorithm optimization of screening code which combines the genetic algorithm and discrete-ordinate method was developed to minimize the costs, size, weight, and so on. This work has some practical significance for gaining the optimization design of shielding. (authors)

Performances of Kevlar and Polyethylene as radiation shielding on-board the International Space Station in high latitude radiation environment.

Science.gov (United States)

Narici, Livio; Casolino, Marco; Di Fino, Luca; Larosa, Marianna; Picozza, Piergiorgio; Rizzo, Alessandro; Zaconte, Veronica

2017-05-10

Passive radiation shielding is a mandatory element in the design of an integrated solution to mitigate the effects of radiation during long deep space voyages for human exploration. Understanding and exploiting the characteristics of materials suitable for radiation shielding in space flights is, therefore, of primary importance. We present here the results of the first space-test on Kevlar and Polyethylene radiation shielding capabilities including direct measurements of the background baseline (no shield). Measurements are performed on-board of the International Space Station (Columbus modulus) during the ALTEA-shield ESA sponsored program. For the first time the shielding capability of such materials has been tested in a radiation environment similar to the deep-space one, thanks to the feature of the ALTEA system, which allows to select only high latitude orbital tracts of the International Space Station. Polyethylene is widely used for radiation shielding in space and therefore it is an excellent benchmark material to be used in comparative investigations. In this work we show that Kevlar has radiation shielding performances comparable to the Polyethylene ones, reaching a dose rate reduction of 32 ± 2% and a dose equivalent rate reduction of 55 ± 4% (for a shield of 10 g/cm 2 ).

Female gonadal shielding with automatic exposure control increases radiation risks

Energy Technology Data Exchange (ETDEWEB)

Kaplan, Summer L.; Zhu, Xiaowei [Children' s Hospital of Philadelphia, Department of Radiology, Philadelphia, PA (United States); University of Pennsylvania, Perelman School of Medicine, Philadelphia, PA (United States); Magill, Dennise; Felice, Marc A. [University of Pennsylvania, Environmental Health and Radiation Safety, Philadelphia, PA (United States); Xiao, Rui [University of Pennsylvania, Department of Biostatistics and Epidemiology, Philadelphia, PA (United States); Ali, Sayed [Temple University Hospital, Department of Radiology, Philadelphia, PA (United States)

2018-02-15

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

Female gonadal shielding with automatic exposure control increases radiation risks

International Nuclear Information System (INIS)

Kaplan, Summer L.; Zhu, Xiaowei; Magill, Dennise; Felice, Marc A.; Xiao, Rui; Ali, Sayed

2018-01-01

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

Female gonadal shielding with automatic exposure control increases radiation risks.

Science.gov (United States)

Kaplan, Summer L; Magill, Dennise; Felice, Marc A; Xiao, Rui; Ali, Sayed; Zhu, Xiaowei

2018-02-01

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation.

Investigation of Radiation Protection Methodologies for Radiation Therapy Shielding Using Monte Carlo Simulation and Measurement

Science.gov (United States)

Tanny, Sean

The advent of high-energy linear accelerators for dedicated medical use in the 1950's by Henry Kaplan and the Stanford University physics department began a revolution in radiation oncology. Today, linear accelerators are the standard of care for modern radiation therapy and can generate high-energy beams that can produce tens of Gy per minute at isocenter. This creates a need for a large amount of shielding material to properly protect members of the public and hospital staff. Standardized vault designs and guidance on shielding properties of various materials are provided by the National Council on Radiation Protection (NCRP) Report 151. However, physicists are seeking ways to minimize the footprint and volume of shielding material needed which leads to the use of non-standard vault configurations and less-studied materials, such as high-density concrete. The University of Toledo Dana Cancer Center has utilized both of these methods to minimize the cost and spatial footprint of the requisite radiation shielding. To ensure a safe work environment, computer simulations were performed to verify the attenuation properties and shielding workloads produced by a variety of situations where standard recommendations and guidance documents were insufficient. This project studies two areas of concern that are not addressed by NCRP 151, the radiation shielding workload for the vault door with a non-standard design, and the attenuation properties of high-density concrete for both photon and neutron radiation. Simulations have been performed using a Monte-Carlo code produced by the Los Alamos National Lab (LANL), Monte Carlo Neutrons, Photons 5 (MCNP5). Measurements have been performed using a shielding test port designed into the maze of the Varian Edge treatment vault.

Radiation shielding for 250 MeV protons

International Nuclear Information System (INIS)

Awschalom, M.

1987-01-01

This paper is targetted at personnel who have the responsibility of designing the radiation shielding against neutron fluences created when protons interact with matter. Shielding of walls and roofs are discussed, as well as neutron dose leakage through labyrinths. Experimental data on neutron flux attenuation are considered, as well as some calculations using the intranuclear cascade calculations and parameterizations

Radiation shielding phenolic fibers and method of producing same

International Nuclear Information System (INIS)

Ohtomo, K.

1976-01-01

A radiation shielding phenolic fiber is described comprising a filamentary phenolic polymer consisting predominantly of a sulfonic acid group-containing cured novolak resin and a metallic atom having a great radiation shielding capacity, the metallic atom being incorporated in the polymer by being chemically bound in the ionic state in the novolak resin. A method for the production of the fiber is discussed

Multifunctional BHL Radiation Shield, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Advances in radiation shielding technology remain an important challenge for NASA in order to protect their astronauts, particularly as NASA grows closer to manned...

Two-dimensional radiation shielding optimization analysis of spent fuel transport container

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

Radiation shielding method for pipes, etc

International Nuclear Information System (INIS)

Nagao, Tetsuya; Takahashi, Shuichi.

1988-01-01

Purpose: To constitute shielding walls of a dense structure around pipes and enable to reduce the wall thickness thereof upon periodical inspection, etc. for nuclear power plants. Constitution: For those portions of pipes requring shieldings, cylindrical vessels surrounding the portions are disposed and connected to a mercury supply system, a mercury discharge system and a freezing system for solidifying mercury. After charging mercury in a tank by way of a supply hose to the cylindrical vessels, the temperature of the mercury is lowered below the freezing point thereof to solidify the mercury while circulating cooling medium, to thereby form dense cylindrical radioactive-ray shielding walls. The specific gravity of mercury is greater than that of lead and, accordingly, the thickness of the shielding walls can be reduced as compared with the conventional wall thickness of the entire laminates. (Takahashi, M.)

Concrete Shielding For Radiation Safety And Unexpected Dangerous Inside Cobalt-60 Industrial Irradiator

International Nuclear Information System (INIS)

Keshk, A.B.; Aly, R.A.

2011-01-01

The study shows a proposed destruction inside one of three cobalt-60 industrial irradiators to determine and reduce the negative results, to improve and modify emergency plan to face terrorism works. The results show the performance of concrete shielding (walls and ceiling) contains the bad effect of dynamic pressures. The explosion forces are prevented to destructive by performance of their concrete shielding, which will contain the most components of devastated systems inside each irradiator after explosion. Shield penetration like electrical cable tunnels, pushers holes, hole with removable plug, product boxes openings, lens opening and ozone duct are affected badly by destruction. Through probability of transporting, some of devastated parts of broken radioactive cobalt- 60 pencils from inside radiation concreter room to outside (surrounded environment) are maintained and causing very danger radiation exposure by gamma rays outside irradiator. A necessity needs to modify emergency plan to prevent any explosive materials to enter inside the main building (irradiation sale) and also discovering any explosive materials which are placed inside the product boxes before passing to inside irradiator. The minimizing radiation exposure (2 mrem/h) inside underground radiation shelters are maintained by reducing radiation dose exerted from a nuclear explosion of 20 kT about 1 km away to a safe value, and calculating the protective factors of radiation main building basements are more than 40 (safety factor) as they are located under ground level, are surrounded by sandy soil and are constructed by concrete. The study shows the proposed basements of the main building maintain success to use as under ground safe radiation shelter (during emergency) with separate safe radiation trace. It begins from the main opening of irradiation sale and leads to underground proposed shelter through modified main stair

Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

International Nuclear Information System (INIS)

Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

1976-01-01

The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

Theoretical analysis of infrared radiation shields of spacecraft

Science.gov (United States)

Shealy, D. L.

1984-01-01

For a system of N diffuse, gray body radiation shields which view only adjacent surfaces and space, the net radiation method for enclosures has been used to formulate a system of linear, nonhomogeneous equations in terms of the temperatures to the fourth power of each surface in the coupled system of enclosures. The coefficients of the unknown temperatures in the system of equations are expressed in terms of configuration factors between adjacent surfaces and the emissivities. As an application, a system of four conical radiation shields for a spin stabilized STARPROBE spacecraft has been designed and analyzed with respect to variations of the cone half angles, the intershield spacings, and emissivities.

Radiation dose to technologists per nuclear medicine examination and estimation of annual dose.

Science.gov (United States)

Bayram, Tuncay; Yilmaz, A Hakan; Demir, Mustafa; Sonmez, Bircan

2011-03-01

Conventional diagnostic nuclear medicine applications have been continuously increasing in most nuclear medicine departments in Turkey, but to our knowledge no one has studied the doses to technologists who perform nuclear medicine procedures. Most nuclear medicine laboratories do not have separate control rooms for technologists, who are quite close to the patient during data acquisition. Technologists must therefore stay behind lead shields while performing their task if they are to reduce the radiation dose received. The aim of this study was to determine external radiation doses to technologists during nuclear medicine procedures with and without a lead shield. Another aim was to investigate the occupational annual external radiation doses to Turkish technologists. This study used a Geiger-Müller detector to measure dose rates to technologists at various distances from patients (0.25, 0.50, 1, and 2 m and behind a lead shield) and determined the average time spent by technologists at these distances. Deep-dose equivalents to technologists were obtained. The following conventional nuclear medicine procedures were considered: thyroid scintigraphy performed using (99m)Tc pertechnetate, whole-body bone scanning performed using (99m)Tc-methylene diphosphonate, myocardial perfusion scanning performed using (99m)Tc-methoxyisobutyl isonitrile, and (201)Tl (thallous chloride) and renal scanning performed using (99m)Tc-dimercaptosuccinic acid. The measured deep-dose equivalent to technologists per procedure was within the range of 0.13 ± 0.05 to 0.43 ± 0.17 μSv using a lead shield and 0.21 ± 0.07 to 1.01 ± 0.46 μSv without a lead shield. Also, the annual individual dose to a technologist performing only a particular scintigraphic procedure throughout a year was estimated. For a total of 95 clinical cases (71 patients), effective external radiation doses to technologists were found to be within the permissible levels. This study showed that a 2-mm lead shield

Analyses of the radiation-caused characteristics change in SOI MOSFETs using field shield isolation

International Nuclear Information System (INIS)

Hirano, Yuuichi; Maeda, Shigeru; Fernandez, Warren; Iwamatsu, Toshiaki; Yamaguchi, Yasuo; Maegawa, Shigeto; Nishimura, Tadashi

1999-01-01

Reliability against radiation ia an important issue in silicon on insulator metal oxide semiconductor field effect transistors (SOI MOSFETs) used in satellites and nuclear power plants and so forth which are severely exposed to radiation. Radiation-caused characteristic change related to the isolation-edge in an irradiated environment was analyzed on SOI MOSFETs. Moreover short channel effects for an irradiated environment were investigated by simulations. It was revealed that the leakage current which was observed in local oxidation of silicon (LOCOS) isolated SOI MOSFETs was successfully suppressed by using field shield isolation. Simulated potential indicated that the potential rise at the LOCOS edge can not be seen in the case of field shield isolation edge which does not have physical isolation. Also it was found that the threshold voltage shift caused by radiation in short channel regime is severer than that in long regime channel. In transistors with a channel length of 0.18μm, a potential rise of the body region by radiation-induced trapped holes can be seen in comparison with that of 1.0μm. As a result, we must consider these effects for designing deep submicron devices used in an irradiated environment. (author)

Important aspects of radiation shielding for fusion reactor tokamaks

International Nuclear Information System (INIS)

Abdou, M.A.

1977-01-01

Radiation shielding is a key subsystem in tokamak reactors. Design of this shield must evolve from economic and technological trade-off studies that account for the strong interrelations among the various components of the reactor system. These trade-offs are examined for the bulk shield on the inner side of the torus and for the special shields of major penetrations. Results derived are applicable for a large class of tokamak-type reactors

Radiation attenuation by lead and nonlead materials used in radiation shielding garments

International Nuclear Information System (INIS)

McCaffrey, J. P.; Shen, H.; Downton, B.; Mainegra-Hing, E.

2007-01-01

The attenuating properties of several types of lead (Pb)-based and non-Pb radiation shielding materials were studied and a correlation was made of radiation attenuation, materials properties, calculated spectra and ambient dose equivalent. Utilizing the well-characterized x-ray and gamma ray beams at the National Research Council of Canada, air kerma measurements were used to compare a variety of commercial and pre-commercial radiation shielding materials over mean energy ranges from 39 to 205 keV. The EGSnrc Monte Carlo user code cavity.cpp was extended to provide computed spectra for a variety of elements that have been used as a replacement for Pb in radiation shielding garments. Computed air kerma values were compared with experimental values and with the SRS-30 catalogue of diagnostic spectra available through the Institute of Physics and Engineering in Medicine Report 78. In addition to garment materials, measurements also included pure Pb sheets, allowing direct comparisons to the common industry standards of 0.25 and 0.5 mm 'lead equivalent'. The parameter 'lead equivalent' is misleading, since photon attenuation properties for all materials (including Pb) vary significantly over the energy spectrum, with the largest variations occurring in the diagnostic imaging range. Furthermore, air kerma measurements are typically made to determine attenuation properties without reference to the measures of biological damage such as ambient dose equivalent, which also vary significantly with air kerma over the diagnostic imaging energy range. A single material or combination cannot provide optimum shielding for all energy ranges. However, appropriate choice of materials for a particular energy range can offer significantly improved shielding per unit mass over traditional Pb-based materials

Shielding technology for high energy radiation production facility

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Heon Il

2004-06-01

In order to develop shielding technology for high energy radiation production facility, references and data for high energy neutron shielding are searched and collected, and calculations to obtain the characteristics of neutron shield materials are performed. For the evaluation of characteristics of neutron shield material, it is chosen not only general shield materials such as concrete, polyethylene, etc., but also KAERI developed neutron shields of High Density PolyEthylene (HDPE) mixed with boron compound (B 2 O 3 , H 2 BO 3 , Borax). Neutron attenuation coefficients for these materials are obtained for later use in shielding design. The effect of source shape and source angular distribution on the shielding characteristics for several shield materials is examined. This effect can contribute to create shielding concept in case of no detail source information. It is also evaluated the effect of the arrangement of shield materials using current shield materials. With these results, conceptual shielding design for PET cyclotron is performed. The shielding composite using HDPE and concrete is selected to meet the target dose rate outside the composite, and the dose evaluation is performed by configuring the facility room conceptually. From the result, the proper shield configuration for this PET cyclotron is proposed

Radiation shielding design for DECY-13 cyclotron using Monte Carlo method

International Nuclear Information System (INIS)

Rasito T; Bunawas; Taufik; Sunardi; Hari Suryanto

2016-01-01

DECY-13 is a 13 MeV proton cyclotron with target H_2"1"8O. The bombarding of 13 MeV protons on target H_2"1"8O produce large amounts of neutrons and gamma radiation. It needs the efficient radiation shielding to reduce the level of neutrons and gamma rays to ensure safety for workers and public. Modeling and calculations have been carried out using Monte Carlo method with MCNPX code to optimize the thickness for the radiation shielding. The calculations were done for radiation shielding of rectangular space room type with the size of 5.5 m x 5 m x 3 m and thickness of 170 cm made from lightweight concrete types of portland. It was shown that with this shielding the dose rate outside the wall was reduced to 1 μSv/h. (author)

Evaluating shielding effectiveness for reducing space radiation cancer risks

International Nuclear Information System (INIS)

Cucinotta, Francis A.; Kim, Myung-Hee Y.; Ren, Lei

2006-01-01

We discuss calculations of probability distribution functions (PDF) representing uncertainties in projecting fatal cancer risk from galactic cosmic rays (GCR) and solar particle events (SPE). The PDFs are used in significance tests for evaluating the effectiveness of potential radiation shielding approaches. Uncertainties in risk coefficients determined from epidemiology data, dose and dose-rate reduction factors, quality factors, and physics models of radiation environments are considered in models of cancer risk PDFs. Competing mortality risks and functional correlations in radiation quality factor uncertainties are included in the calculations. We show that the cancer risk uncertainty, defined as the ratio of the upper value of 95% confidence interval (CI) to the point estimate is about 4-fold for lunar and Mars mission risk projections. For short-stay lunar missions ( 180d) or Mars missions, GCR risks may exceed radiation risk limits that are based on acceptable levels of risk. For example, the upper 95% CI exceeding 10% fatal risk for males and females on a Mars mission. For reducing GCR cancer risks, shielding materials are marginally effective because of the penetrating nature of GCR and secondary radiation produced in tissue by relativistic particles. At the present time, polyethylene or carbon composite shielding cannot be shown to significantly reduce risk compared to aluminum shielding based on a significance test that accounts for radiobiology uncertainties in GCR risk projection

Effectiveness of construction materials and some minerals used as radiation shielding

International Nuclear Information System (INIS)

Khunarak, P.; Bunnak, S.

1988-01-01

There are many kinds of ores in Thailand, some large amount of them are cheap and easy to obtain possess shielding properties for gamma radiation. These ores are baryte, illmenite, galena, scheelite, wolframite pyrite, cerrusite. Besides, building structure materials are also introduced for shielding properties study by using Co-60, Cs-137 and Ra-226 as gamma radiation sources in the experiments. The results turn out that those high density ores will possess a better shielding property than the low density ores. Radiation measurement equipment is G.M. tube connected to rate meter

Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

International Nuclear Information System (INIS)

Viswanathan, S.; Jose, M.T.; Venkatraman, B.

2016-01-01

In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

Shielding Benchmark Computational Analysis

International Nuclear Information System (INIS)

Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

2000-01-01

Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

Barium-borate-flyash glasses: As radiation shielding materials

International Nuclear Information System (INIS)

Singh, Sukhpal; Kumar, Ashok; Singh, Devinder; Thind, Kulwant Singh; Mudahar, Gurmel S.

2008-01-01

The attenuation coefficients of barium-borate-flyash glasses have been measured for γ-ray photon energies of 356, 662, 1173 and 1332 keV using narrow beam transmission geometry. The photon beam was highly collimated and overall scatter acceptance angle was less than 3 o . Our results have an uncertainty of less than 3%. These coefficients were then used to obtain the values of mean free path (mfp), effective atomic number and electron density. Good agreements have been observed between experimental and theoretical values of these parameters. From the studies of the obtained results it is reported here that from the shielding point of view the barium-borate-flyash glasses are better shields to γ-radiations in comparison to the standard radiation shielding concretes and also to the ordinary barium-borate glasses

Development of highly effective neutron shielding material made of phenol-novolac type epoxy resin

Energy Technology Data Exchange (ETDEWEB)

Cho, Soo Haeng; Jeong, Myeong Soo; Hong, Sun Seok; Lee, Won Kyoung; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy; Oh, Seok Jin

1998-06-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. Especially we developed phenol-novolac type epoxy resin based neutron shielding materials and their characteristics were also evaluated. (author). 22 refs., 11 tabs., 21 figs

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

Energy Technology Data Exchange (ETDEWEB)

Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok [Nonproliferation System Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Chang Je [Dept. of Nuclear Engineering, Sejong University, Seoul (Korea, Republic of)

2015-04-15

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101{sup 2n}/cm{sup 2}·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B{sub 4}C, and Li{sub 2}CO{sub 3}] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

International Nuclear Information System (INIS)

Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok; Park, Chang Je

2015-01-01

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101 2n /cm 2 ·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B 4 C, and Li 2 CO 3 ] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

Directory of Open Access Journals (Sweden)

Jeong Dong Kim

2015-04-01

Full Text Available A lead slowing-down spectrometer (LSDS system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea is planned to utilize a high-flux (>1012 n/cm2·s neutron source comprised of a high-energy (30 MeV/high-current (∼2 A electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h, a few shielding materials [high-density polyethylene (HDPE–Borax, B4C, and Li2CO3] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near

Shielding Factors for Gamma Radiation from Activity Deposited on Structures and Ground Surfaces

DEFF Research Database (Denmark)

Jensen, Per Hedemann

1985-01-01

A computer model DEPSHIELD for the calculation of shielding factors for gamma radiation at indoor residences in multistorey and single-family houses has been developed. The model is based on the exponential point kernel that links the radiation flux density at a given detector point to a point...... it possible to determine the dose reduction effect from a decontamination of the different surfaces. The model has been used in a study of the consequences of land contamination of Danish territory after hypothetical core-melt accidents at the Barseback nuclear power plant in Sweden. The model has also been...

Gamma radiation shielding analysis of lead-flyash concretes

International Nuclear Information System (INIS)

Singh, Kanwaldeep; Singh, Sukhpal; Dhaliwal, A.S.; Singh, Gurmel

2015-01-01

Six samples of lead-flyash concrete were prepared with lead as an admixture and by varying flyash content – 0%, 20%, 30%, 40%, 50% and 60% (by weight) by replacing cement and keeping constant w/c ratio. Different gamma radiation interaction parameters used for radiation shielding design were computed theoretically and measured experimentally at 662 keV, 1173 keV and 1332 keV gamma radiation energy using narrow transmission geometry. The obtained results were compared with ordinary-flyash concretes. The radiation exposure rate of gamma radiation sources used was determined with and without lead-flyash concretes. - Highlights: • Concrete samples with lead as admixture were casted with flyash replacing 0%, 20%, 30%, 40%, 50% and 60% of cement content (by weight). • Gamma radiation shielding parameters of concretes for different gamma ray sources were measured. • The attenuation results of lead-flyash concretes were compared with the results of ordinary flyash concretes

Nuclear steam generator tubesheet shield

International Nuclear Information System (INIS)

Nickerson, J.H.D.; Ruhe, A.

1982-01-01

The invention involves improvements to a nuclear steam generator of the type in which a plurality of U-shaped tubes are connected at opposite ends to a tubesheet and extend between inlet and outlet chambers, with the steam generator including an integral preheater zone adjacent to the downflow legs of the U-shaped tubes. The improvement is a thermal shield disposed adjacent to an upper face of the tubesheet within the preheater zone, the shield including ductile cladding material applied directly to the upper face of the tubesheet, with the downflow legs of the U-shaped tubes extending through the cladding into the tubesheet

Shielding ability of lead loaded radiation resistant gloves

International Nuclear Information System (INIS)

Kawano, Takao; Ebihara, Hiroshi

1990-01-01

The shielding ability of radiation resistant gloves were examined. The gloves are made of lead loaded (as PbO 2 ) polyvinyl chloride resin and are about 0.4 mm of thickness (70 mg/cm 2 ). Eleven test pieces were sampled from each of three gloves (total were thirty three) and the transmission rates for radiations (X-ray or γ-ray) through the test pieces were measured with radiation sources, 99m Tc, 57 Co, 133 Ba, 133 Xe and 241 Am. The differences of the transmission rate for radiations by the positions of the gloves were smaller than 15%, and the differences by three gloves were smaller than 5% in the case of 60 keV and 141 keV radiations. The average transmission rates for radiations in thirty three test pieces were about 40% for 30 keV radiation, about 90% for 80 keV and 140 keV radiations. The shielding characteristic of the gloves could be equivalent to about 0.026 mm thick lead plate. (author)

Radiation shielding structure for concrete structure

International Nuclear Information System (INIS)

Oya, Hiroshi

1998-01-01

Crack inducing members for inducing cracks in a predetermined manner are buried in a concrete structure. Namely, a crack-inducing member comprises integrally a shielding plate and extended plates situated at the center of a wall and inducing plates vertically disposed to the boundary portion between them with the inducing plates being disposed each in a direction perforating the wall. There are disposed integrally a pair of the inducing plate spaced at a predetermined horizontal distance on both sides of the shielding plate so as to form a substantially crank-shaped cross section and extended plates formed in the extending direction of the shielding plate, and the inducing plates are disposed each in a direction perforating the wall. Then, cracks generated when stresses are exerted can be controlled, and generation of cracks passing through the concrete structure can be prevented reliably. The reliability of a radiation shielding effect can be enhanced remarkably. (N.H.)

Evaluation of the performance of peridotite aggregates for radiation shielding concrete

International Nuclear Information System (INIS)

Wang, Jinjun; Li, Guofeng; Meng, Dechuan

2014-01-01

Highlights: • Using peridotite rich in crystal water as aggregates of radiation-shielding concrete. • Performance of peridotite concrete is simulated and compared with ordinary concrete. • Performance of concrete samples is tested. • Neutron shielding performance can be significantly enhanced by peridotite aggregates. - Abstract: Peridotite is a kind of material that is rich in crystal water. In this paper, peridotite is used as fine and coarse aggregates for radiation shielding concrete. The transmission data of different concrete thickness and different energy neutron are calculated using Monte-Carlo method. The neutron shielding performance of the peridotite concrete samples are tested using 241 Am-Be neutron source. The results show that the peridotite is an excellent neutron shielding material

Beta Bremsstrahlung dose in concrete shielding

Energy Technology Data Exchange (ETDEWEB)

Manjunatha, H.C., E-mail: manjunatha@rediffmail.com [Department of Physics, Government college for women, Kolar 563101, Karnataka (India); Chandrika, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India); Rudraswamy, B. [Department of Physics, Bangalore University, Bangalore 560056, Karnataka (India); Sankarshan, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India)

2012-05-11

In a nuclear reactor, beta nuclides are released during nuclear reactions. These betas interact with shielding concrete and produces external Bremsstrahlung (EB) radiation. To estimate Bremsstrahlung dose and shield efficiency in concrete, it is essential to know Bremsstrahlung distribution or spectra. The present work formulated a new method to evaluate the EB spectrum and hence Bremsstrahlung dose of beta nuclides ({sup 32}P, {sup 89}Sr, {sup 90}Sr-{sup 90}Y, {sup 90}Y, {sup 91}Y, {sup 208}Tl, {sup 210}Bi, {sup 234}Pa and {sup 40}K) in concrete. The Bremsstrahlung yield of these beta nuclides in concrete is also estimated. The Bremsstrahlung yield in concrete due to {sup 90}Sr-{sup 90}Y is higher than those of other given nuclides. This estimated spectrum is accurate because it is based on more accurate modified atomic number (Z{sub mod}) and Seltzer's data, where an electron-electron interaction is also included. Presented data in concrete provide a quick and convenient reference for radiation protection. The present methodology can be used to calculate the Bremsstrahlung dose in nuclear shielding materials. It can be quickly employed to give a first pass dose estimate prior to a more detailed experimental study. - Highlights: Black-Right-Pointing-Pointer Betas released in a nuclear reactor interact with shielding concrete and produces Bremsstrahlung. Black-Right-Pointing-Pointer The present work formulated a new method to evaluate the Bremsstrahlung spectrum and dose in concrete. Black-Right-Pointing-Pointer Presented data in concrete provide a quick and convenient reference for radiation protection.

Radiation, ionization, and detection in nuclear medicine

International Nuclear Information System (INIS)

Gupta, Tapan K.

2013-01-01

Up-to-date information on a wide range of topics relating to radiation, ionization, and detection in nuclear medicine. In-depth coverage of basic radiophysics relating to diagnosis and therapy. Extensive discussion of instrumentation and radiation detectors. Detailed information on mathematical modelling of radiation detectors. Although our understanding of cancer has improved, the disease continues to be a leading cause of death across the world. The good news is that the recent technological developments in radiotherapy, radionuclide diagnostics and therapy, digital imaging systems, and detection technology have raised hope that cancer will in the future be combatted more efficiently and effectively. For this goal to be achieved, however, safe use of radionuclides and detailed knowledge of radiation sources are essential. Radiation, Ionization, and Detection in Nuclear Medicine addresses these subjects and related issues very clearly and elaborately and will serve as the definitive source of detailed information in the field. Individual chapters cover fundamental aspects of nuclear radiation, including dose and energy, sources, and shielding; the detection and measurement of radiation exposure, with detailed information on mathematical modelling; medical imaging; the different types of radiation detector and their working principles; basic principles of and experimental techniques for deposition of scintillating materials; device fabrication; the optical and electrical behaviors of radiation detectors; and the instrumentation used in nuclear medicine and its application. The book will be an invaluable source of information for academia, industry, practitioners, and researchers.

Radiation, ionization, and detection in nuclear medicine

Energy Technology Data Exchange (ETDEWEB)

Gupta, Tapan K. [Radiation Monitoring Devices Research, Nuclear Medicine, Watertown, MA (United States)

2013-08-01

Up-to-date information on a wide range of topics relating to radiation, ionization, and detection in nuclear medicine. In-depth coverage of basic radiophysics relating to diagnosis and therapy. Extensive discussion of instrumentation and radiation detectors. Detailed information on mathematical modelling of radiation detectors. Although our understanding of cancer has improved, the disease continues to be a leading cause of death across the world. The good news is that the recent technological developments in radiotherapy, radionuclide diagnostics and therapy, digital imaging systems, and detection technology have raised hope that cancer will in the future be combatted more efficiently and effectively. For this goal to be achieved, however, safe use of radionuclides and detailed knowledge of radiation sources are essential. Radiation, Ionization, and Detection in Nuclear Medicine addresses these subjects and related issues very clearly and elaborately and will serve as the definitive source of detailed information in the field. Individual chapters cover fundamental aspects of nuclear radiation, including dose and energy, sources, and shielding; the detection and measurement of radiation exposure, with detailed information on mathematical modelling; medical imaging; the different types of radiation detector and their working principles; basic principles of and experimental techniques for deposition of scintillating materials; device fabrication; the optical and electrical behaviors of radiation detectors; and the instrumentation used in nuclear medicine and its application. The book will be an invaluable source of information for academia, industry, practitioners, and researchers.

Radiation Shielding of Lunar Regolith/Polyethylene Composites and Lunar Regolith/Water Mixtures

Science.gov (United States)

Johnson, Quincy F.; Gersey, Brad; Wilkins, Richard; Zhou, Jianren

2011-01-01

Space radiation is a complex mixed field of ionizing radiation that can pose hazardous risks to sophisticated electronics and humans. Mission planning for lunar exploration and long duration habitat construction will face tremendous challenges of shielding against various types of space radiation in an attempt to minimize the detrimental effects it may have on materials, electronics, and humans. In late 2009, the Lunar Crater Observation and Sensing Satellite (LCROSS) discovered that water content in lunar regolith found in certain areas on the moon can be up to 5.6 +/-2.8 weight percent (wt%) [A. Colaprete, et. al., Science, Vol. 330, 463 (2010). ]. In this work, shielding studies were performed utilizing ultra high molecular weight polyethylene (UHMWPE) and aluminum, both being standard space shielding materials, simulated lunar regolith/ polyethylene composites, and simulated lunar regolith mixed with UHMWPE particles and water. Based on the LCROSS findings, radiation shielding experiments were conducted to test for shielding efficiency of regolith/UHMWPE/water mixtures with various percentages of water to compare relative shielding characteristics of these materials. One set of radiation studies were performed using the proton synchrotron at the Loma Linda Medical University where high energy protons similar to those found on the surface of the moon can be generated. A similar experimental protocol was also used at a high energy spalation neutron source at Los Alamos Neutron Science Center (LANSCE). These experiments studied the shielding efficiency against secondary neutrons, another major component of space radiation field. In both the proton and neutron studies, shielding efficiency was determined by utilizing a tissue equivalent proportional counter (TEPC) behind various thicknesses of shielding composite panels or mixture materials. Preliminary results from these studies indicated that adding 2 wt% water to regolith particles could increase shielding of

Shield nuclear design for the 5-kWe TE system

International Nuclear Information System (INIS)

Keshishian, V.

1972-01-01

The nuclear analysis of the 5-kW(e) reactor shield is presented. Calculation methods and optimization techniques used are presented. Borated stainless steel was selected for the gamma ray shield with tungsten alloy as an alternate. The total shield weight was calculated to be 355 lb. (U.S.)

Radiation shielding aspects for long manned mission to space: Criteria, survey study, and preliminary model

OpenAIRE

Sztejnberg Manuel; Xiao Shanjie; Satvat Nader; Limón Felisa; Hopkins John; Jevremović Tatjana

2006-01-01

The prospect of manned space missions outside Earth's orbit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is therefore obvious that nuclear power production in space is a must. On th...

Temperature distribution due to the heat generation in nuclear reactor shielding

International Nuclear Information System (INIS)

Torres, L.M.R.

1985-01-01

A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

Radioactivity, shielding, radiation damage, and remote handling

International Nuclear Information System (INIS)

Wilson, M.T.

1975-01-01

Proton beams of a few hundred million electron volts of energy are capable of inducing hundreds of curies of activity per microampere of beam intensity into the materials they intercept. This adds a new dimension to the parameters that must be considered when designing and operating a high-intensity accelerator facility. Large investments must be made in shielding. The shielding itself may become activated and require special considerations as to its composition, location, and method of handling. Equipment must be designed to withstand large radiation dosages. Items such as vacuum seals, water tubing, and electrical insulation must be fabricated from radiation-resistant materials. Methods of maintaining and replacing equipment are required that limit the radiation dosages to workers.The high-intensity facilities of LAMPF, SIN, and TRIUMF and the high-energy facility of FERMILAB have each evolved a philosophy of radiation handling that matches their particular machine and physical plant layouts. Special tooling, commercial manipulator systems, remote viewing, and other techniques of the hot cell and fission reactor realms are finding application within accelerator facilities. (U.S.)

Using natural local materials for developing special radiation shielding concretes, and deduction of its shielding characteristics

International Nuclear Information System (INIS)

Kharita, M. H.; Takeyeddin, M.; Al-Nassar, M.; Yousef, S.

2006-06-01

Concrete is considered as the most important material to be used for radiation shielding in facilities contain radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the construction of the concrete, which is highly relative to the composing aggregates i.e. aggregates consist about 70 - 80% of the total weight of normal concrete. In this project tow types of concrete used in Syria (in Damascus and Aleppo) had been studied and their shielding properties were defined for gamma ray from Cs-137 and Co-60 sources, and for neutrons from Am-Be source. About 10% reduction in HVL was found in the comparison between the tow concrete types for both neutrons and gammas. Some other types of concrete were studied using aggregates from different regions in Syria, to improve the shielding properties of concrete, and another 10% of reduction was achieved in comparison with Damascene concrete (20% in comparison with the concrete from Aleppo) for both neutrons and gamma rays. (author)

Analysis of radiation shields of BNPP spent fuel pool

International Nuclear Information System (INIS)

Ayoobian, N.; Hadad, K.; Nematollahi, M. R.

2007-01-01

Radioactive protection is one of the most important subjects in nuclear power plants safety. Analysis of BNPP spent fuel pool shielding , as a main source of energetic γ-rays was the main goal of this project. Firstly, we simulated the reactor core using WIMSD-4 neutronic code and the amount of fission product in the fuel assembly (FA) was calculated during the reactor operation. Then, by obtaining the results from the previous calculation and by using MCNP4C nuclear code , the intensity of γ-rays was obtained in layers of spent fuel pool shields. The results have shown that no significant γ-rays passed through these shields. Finally, an accident and resulting exposure dose above the pool was analyzed

Development of radiation shielding materials and NBC pads for infantry combat vehicle and tank

International Nuclear Information System (INIS)

Pal, R.S.; Gautam, O.P.; Katiyar, Mohit; Tripathi, D.N.; Singh, R.K.

2008-01-01

Tanks have special lining materials inside, providing a certain degree of radiation protection for operation in nuclear scenario's. At present these special lining materials in the form of sheets are imported and are fitted into armoured vehicles. Three types of polymer compositions; PE(M)SE, PEC-ISE and PEC-IISE were formulated based on polymer matrix, specific fillers and anti-ageing additives. Prototype NBC pads based on polymer composition PEC-ISE was finalized for moulding of NBC pads for use in ICVs and composition PE(M)SE was finalized for T-90 tanks. The physico-mechanical properties for NBC pads have been evaluated. Radiography of test samples was conducted to ensure homogeneity of specific fillers in the polymer matrix. Radiation shielding factors against nuclear radiation sources ( 60 Co, I37 Cs and 252 Cf) were evaluated at DL Jodhpur and found to be better than imported Russian Pads designed for ICVs and T-90 Tanks. Drawings for twelve types of NBC pads for ICVs and one hundred eighteen types of pads for T-90 tank were generated with the help of design tool, Auto Desk Inventor-II and metallic moulds for moulding of NBC pads were fabricated. Prototype NBC pads were moulded through compression moulding process. Radiation protection factors of prototype NBC pads, after fitment in ICVs, were also evaluated against neutron and gamma (primary and secondary) radiation sources. Prototype NBC pads for ICVs have shown 20% improvement in overall protection level and NBC pads for T-90 tanks have been developed as per design requirements. Manufacturing facility for NBC shielding pads have been established in association with industries. (author)

[The model of radiation shielding of the service module of the International space station].

Science.gov (United States)

Kolomenskiĭ, A V; Kuznetsov, V G; Laĭko, Iu A; Bengin, V V; Shurshakov, V A

2001-01-01

Compared and contrasted were models of radiation shielding of habitable compartments of the basal Mir module that had been used to calculate crew absorbed doses from space radiation. Developed was a model of the ISS Service module radiation shielding. It was stated that there is a good agreement between experimental shielding function and the one calculated from this model.

Shield or not to Shield: Effects of Solar Radiation on Water Temperature Sensor Accuracy

Directory of Open Access Journals (Sweden)

Robert L. Wilby

2013-10-01

Full Text Available Temperature sensors are potentially susceptible to errors due to heating by solar radiation. Although this is well known for air temperature (Ta, significance to continuous water temperature (Tw monitoring is relatively untested. This paper assesses radiative errors by comparing measurements of exposed and shielded Tinytag sensors under indirect and direct solar radiation, and in laboratory experiments under controlled, artificial light. In shallow, still-water and under direct solar radiation, measurement discrepancies between exposed and shielded sensors averaged 0.4 °C but can reach 1.6 °C. Around 0.3 °C of this inconsistency is explained by variance in measurement accuracy between sensors; the remainder is attributed to solar radiation. Discrepancies were found to increase with light intensity, but to attain Tw differences in excess of 0.5 °C requires direct, bright solar radiation (>400 W m−2 in the total spectrum. Under laboratory conditions, radiative errors are an order of magnitude lower when thermistors are placed in flowing water (even at velocities as low as 0.1 m s−1. Radiative errors were also modest relative to the discrepancy between different thermistor manufacturers. Based on these controlled experiments, a set of guidelines are recommended for deploying thermistor arrays in water bodies.

A study on radiation shielding and safety analysis for a synchrotron radiation beamline

International Nuclear Information System (INIS)

Asano, Yoshihiro

2001-03-01

Methods of shielding design and safety analysis are presented for a beam-line of synchrotron radiation. This paper consists of the shielding and safety study of synchrotron radiation with extremely intense and low energy photon below several hundreds keV, and the study for the behavior of remarkable high-energy photons up to 8 GeV, which can creep into beam-lines. A new shielding design code, STAC8 was developed to estimate the leakage dose outside the beam line hutch (an enclosure of the beam, optical elements or experimental instruments) easily and quickly with satisfactory accuracy. The code can calculate consistently from sources of synchrotron radiation to dose equivalent outside hutches with considering the build up effect and polarization effect. Validity of the code was verified by comparing its calculations with those of Monte Carlo simulations and measurement results of the doses inside the hutch of the BL14C of Photon Factory in the High Energy Accelerator Research Organization (KEK), showing good agreements. The shielding design calculations using STAC8 were carried out to apply to the practical beam-lines with the considering polarization effect and clarified the characteristics of the typical beam-line of the third generation synchrotron radiation facility, SPring-8. In addition, the shielding calculations were compared with the measurement outside the shield wall of the bending magnet beam-line of SPring-8, and showed fairly good agreement. The new shielding problems, which have usually been neglected in shielding designs for existing synchrotron radiation facilities, are clarified through the analysis of the beam-line shielding of SPring-8. The synchrotron radiation from the SPring-8 has such extremely high-intensity involving high energy photons that the scattered synchrotron radiation from the concrete floor of the hutch, the ground shine, causes a seriously high dose. The method of effective shielding is presented. For the estimation of the gas

A study on radiation shielding and safety analysis for a synchrotron radiation beamline

Energy Technology Data Exchange (ETDEWEB)

Asano, Yoshihiro [Japan Atomic Energy Research Inst., Kansai Research Establishment, Synchrotron Radiation Research Center, Mikazuhi, Hyogo (Japan)

2001-03-01

Methods of shielding design and safety analysis are presented for a beam-line of synchrotron radiation. This paper consists of the shielding and safety study of synchrotron radiation with extremely intense and low energy photon below several hundreds keV, and the study for the behavior of remarkable high-energy photons up to 8 GeV, which can creep into beam-lines. A new shielding design code, STAC8 was developed to estimate the leakage dose outside the beam line hutch (an enclosure of the beam, optical elements or experimental instruments) easily and quickly with satisfactory accuracy. The code can calculate consistently from sources of synchrotron radiation to dose equivalent outside hutches with considering the build up effect and polarization effect. Validity of the code was verified by comparing its calculations with those of Monte Carlo simulations and measurement results of the doses inside the hutch of the BL14C of Photon Factory in the High Energy Accelerator Research Organization (KEK), showing good agreements. The shielding design calculations using STAC8 were carried out to apply to the practical beam-lines with the considering polarization effect and clarified the characteristics of the typical beam-line of the third generation synchrotron radiation facility, SPring-8. In addition, the shielding calculations were compared with the measurement outside the shield wall of the bending magnet beam-line of SPring-8, and showed fairly good agreement. The new shielding problems, which have usually been neglected in shielding designs for existing synchrotron radiation facilities, are clarified through the analysis of the beam-line shielding of SPring-8. The synchrotron radiation from the SPring-8 has such extremely high-intensity involving high energy photons that the scattered synchrotron radiation from the concrete floor of the hutch, the ground shine, causes a seriously high dose. The method of effective shielding is presented. For the estimation of the gas

Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

Multiconfigurational self-consistent field calculations of nuclear shieldings using London atomic orbitals

DEFF Research Database (Denmark)

Ruud, Kenneth; Helgaker, Trygve; Kobayashi, Rika

1994-01-01

to corresponding individual gauges for localized orbitals (IGLO) results. The London results show better basis set convergence than IGLO, especially for heavier atoms. It is shown that the choice of active space is crucial for determination of accurate nuclear shielding constants.......Nuclear shielding calculations are presented for multiconfigurational self-consistent field wave functions using London atomic orbitals (gauge invariant atomic orbitals). Calculations of nuclear shieldings for eight molecules (H2O, H2S, CH4, N2, CO, HF, F2, and SO2) are presented and compared...

Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

International Nuclear Information System (INIS)

Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

2011-01-01

The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

Radiation shielding technology development for proton linear accelerator

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Ouk; Lee, Y. O.; Cho, Y. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, M. H.; Sin, M. W.; Park, B. I. [Kyunghee Univ., Seoul (Korea, Republic of)] [and others

2005-09-01

This report was presented as an output of 2-year project of the first phase Proton Engineering Frontier Project(PEFP) on 'Radiation Shielding Technology Development for Proton Linear Accelerator' for 20/100 MeV accelerator beam line and facility. It describes a general design concept, provision and update of basic design data, and establishment of computer code system. It also includes results of conceptual and preliminary designs of beam line, beam dump and beam facilities as well as an analysis of air-activation inside the accelerator equipment. This report will guides the detailed shielding design and production of radiation safety analysis report scheduled in the second phase project.

Shielding practice

International Nuclear Information System (INIS)

Sauermann, P.F.

1985-08-01

The basis of shielding practice against external irradiation is shown in a simple way. For most sources of radiation (point sources) occurring in shielding practice, the basic data are given, mainly in the form of tables, which are required to solve the shielding problems. The application of these data is explained and discussed using practical examples. Thickness of shielding panes of glove boxes for α and β radiation; shielding of sealed γ-radiography sources; shielding of a Co-60 radiation source, and of the manipulator panels for hot cells; damping factors for γ radiation and neutrons; shielding of fast and thermal neutrons, and of bremsstrahlung (X-ray tubes, Kr-85 pressure gas cylinders, 42 MeV betatrons, 20 MeV linacs); two-fold shielding (lead glass windows for hot cells, 14 MeV neutron generators); shielding against scattered radiation. (orig./HP) [de

Radiation shielding with Bi2O3 and ZrO2:Y composites: preparation and characterization

International Nuclear Information System (INIS)

Fontainha, Crissia C.P.

2015-01-01

Despite the benefits of medical imaging examinations, there is a worrying contribution of dose of radiation to population due to the high dose procedures. Procedures as interventional radiology, Computed Tomography (CT) and nuclear medicine provide high doses to the skin of patients, provoking radiation deleterious effects. New attenuators materials have been widely investigated for radiation shielding in those regions of high risk, allowing significant dose reduction near the patient's skin. Composites with Bi 2 O 3 and ZrO 2 :Y metals were obtained by mixing them with P(VDF-TrFe) copolymers from casting. Composites were produced with concentrations of 2, 4 and 8% wt. of Yttrium stabilized zirconia. Bi 2 O 3 containing composites were produced with the same concentrations (2, 4 and 8% wt.), with Bi 2 O 3 particles being previously functionalized with methacrylic acid (MAA). The composites were characterized by FTIR. The entrance skin dose characterization was performed with and without the use of radiation protective shielding. The composite samples were exposed to an absorbed dose of 100 mGy of RQR5 beam quality (70 kV X-ray beam). The attenuation factors, evaluated by XR-QA2 radiochromic films, indicate that both P(VDF-TrFE)/Bi 2 O 3 and P(VDF-TrFE)/ZrO 2 :Y composites are good candidates for use as patient radiation shielding in high dose medical procedures. (author)

Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

International Nuclear Information System (INIS)

Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

1976-01-01

The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

International Nuclear Information System (INIS)

Chen, Y.; Fischer, U.

2005-01-01

Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

Radiation shielding design for a hot repair facility

International Nuclear Information System (INIS)

Courtney, J.C.; Dwight, C.C.

1991-01-01

A new repair and decontamination area is being built to support operations at the demonstration fuel cycle facility for the Integral Fast Reactor program at Argonne National Laboratory's site at the Idaho National Engineering Laboratory. Provisions are made for remote, glove wall, and contact maintenance on equipment removed from hot cells where spent fuel will be electrochemically processed and recycled to the Experimental Breeder Reactor-II. The source for the shielding design is contamination from a mix of fission and activation products present on items removed from the hot cells. The repair facility also serves as a transfer path for radioactive waste produced by processing operations. Radiation shields are designed to limit dose rates to no more than 5 microSv h-1 (0.5 mrem h-1) in normally occupied areas. Point kernel calculations with buildup factors have been used to design the shielding and to position radiation monitors within the area

Gamma radiation shielding materials improved with burning resistance

International Nuclear Information System (INIS)

Nakamura, Michio; Nakamura, Ken-ichi; Yukawa, Katsunori.

1985-01-01

Purpose: To obtain gamma irradiation shielding materials excellent in workability and resistant to burning by using a two component type room temperature vulcanizing silicon rubber composition as the base material. Method: Silicon rubber comprising a diorganopolysiloxane polymer, an alkyl silicate as a crosslinker and a suitable sulfurdizing catalyst, for example, a carboxylate is mixed with iron powder and silicon oxide powder as reinforcing and flame retardant material and applied with molding. The iron powder and the silica rocks powder have grain size of 50 - 150 μm and 1 - 70 μm and charged by the amount of from 55 to 60 % by weight and from 20 to 25 % by weight respectively. The fluidizing property is impaired if the particle size of the silica rocks powder is less than 1 μm and, while on the other hand, no desired specific gravity of a predetermined value can be obtained for the molding product if the filled amount of the iron powder is less than 55 %. The oxygen index of the molding product is 45 to improve the burning resistance. The materials are excellent in the air-tightness, gamma radiation shielding performance, elasticity and workability required for the cable penetrations in a nuclear power plant and they generate noxious gases neither. (Kawakami, Y.)

Can active proton interrogation find shielded nuclear threats at human-safe radiation levels?

Energy Technology Data Exchange (ETDEWEB)

Liew, Seth Van, E-mail: vanliew@gmail.com

2017-05-21

A new method of low-dose proton radiography is presented. The system is composed of an 800 MeV proton source, bending magnets, and compact detectors, and is designed for drive-through cargo scanning. The system has been simulated using GEANT4. Material identification algorithms and pixel sorting methods are presented that allow the system to perform imaging at doses low enough to scan passenger vehicles and people. Results are presented on imaging efficacy of various materials and cluttered cargoes. The identification of shielded nuclear materials at human-safe doses has been demonstrated.

Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study

Data.gov (United States)

National Aeronautics and Space Administration — The objectives of the proposed research are to develop a space radiation shielding material system that has high efficacy for shielding radiation and also has high...

Proposal of secure camera-based radiation warning system for nuclear detection

International Nuclear Information System (INIS)

Tsuchiya, Ken'ichi; Kurosawa, Kenji; Akiba, Norimitsu; Kakuda, Hidetoshi; Imoto, Daisuke; Hirabayashi, Manato; Kuroki, Kenro

2016-01-01

Counter-terrorisms against radiological and nuclear threat are significant issues toward Tokyo 2020 Olympic and Paralympic Games. In terms of cost benefit, it is not easy to build a warning system for nuclear detection to prevent a Dirty Bomb attack (dispersion of radioactive materials using a conventional explosive) or a Silent Source attack (hidden radioactive materials) from occurring. We propose a nuclear detection system using the installed secure cameras. We describe a method to estimate radiation dose from noise pattern in CCD images caused by radiation. Some dosimeters under neutron and gamma-ray irradiations (0.1mSv-100mSv) were taken in CCD video camera. We confirmed amount of noise in CCD images increased in radiation exposure. The radiation detection using CMOS in secure cameras or cell phones has been implemented. However, in this presentation, we propose a warning system including neutron detection to search shielded nuclear materials or radiation exposure devices using criticality. (author)

A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

Hot Cell Window Shielding Analysis Using MCNP

International Nuclear Information System (INIS)

Pope, Chad L.; Scates, Wade W.; Taylor, J. Todd

2009-01-01

The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

Radiation safely culture in nuclear facilities

International Nuclear Information System (INIS)

Coates, R.

2018-01-01

The importance of developing a sound radiation safety culture is a relatively new development in the practical application of radiation protection in operational facilities. It is instructive to trace the evolution of the fundamental approaches to controlling operational exposures, staring with the engineering-based 'Distance, Shielding and Time' mantra, through the growing emphasis on ALARA and systematic management-based approaches, towards a recognition of the importance of developing a more 'hearts and minds' approach based within the wider safety culture of the organization. The underlying requirements for developing a strong radiation safety culture are not novel, and are largely identical to those necessary for nuclear safety culture, which is why an integrated approach to culture within the organization is essential

Radiation shield ring assembly and method of disassembling components of a nuclear steam generator using such assembly

International Nuclear Information System (INIS)

Meuschke, R.E.; Wolfe, D.L.

1982-01-01

This invention relates to an apparatus and a method for cutting, within a shielding confinement, the irradiated components of a nuclear steam generator to reduce such components to a size to permit their subsequent removal from the containment structure of the generator

Radiation shielding calculation for the MOX fuel fabrication plant Melox

International Nuclear Information System (INIS)

Lee, Y.K.; Nimal, J.C.; Chiron, M.

1994-01-01

Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

Analytic Shielding Optimization to Reduce Crew Exposure to Ionizing Radiation Inside Space Vehicles

Science.gov (United States)

Gaza, Razvan; Cooper, Tim P.; Hanzo, Arthur; Hussein, Hesham; Jarvis, Kandy S.; Kimble, Ryan; Lee, Kerry T.; Patel, Chirag; Reddell, Brandon D.; Stoffle, Nicholas;

Dictionary of radiation protection, radiobiology and nuclear medicine. English-German-French-Russian

International Nuclear Information System (INIS)

Sube, R.

1985-01-01

This multilingual dictionary covers the subject fields of radiation protection, radiobiology, and nuclear medicine with about 12,000 terms in each language. All terms are supplemented by one or more abbreviations of 22 special branches to assure the use of the very relevant terms. Special branches listed are for instance decontamination, dosimetry, atomic legislation, radiation detectors, radiography (medical), radiotherapy, safeguards, shielding, tansportation and storage. The terminology used in the International Nuclear Information System (INIS) of the IAEA has been completely taken into account

Development of special radiation shielding concretes using natural local materials and evaluation of their shielding characteristics

International Nuclear Information System (INIS)

Kharita, M. H.; Takeyeddin, M.; Al-Nassri, M.; Yousef, S.

2008-01-01

Concrete is one of the most important materials used for radiation shielding in facilities containing radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the composite of the concrete. Aggregates is the largest constituent (about 70-80% of the total weight of normal concrete). The aim of this work is to develop special concrete with good shielding properties for gamma and neutrons, using natural local materials. For this reason two types of typical concrete widely used in Syria (in Damascus and Aleppo) and four other types of concrete, using aggregates from different regions, have been prepared. The shielding properties of these six types were studied for gamma ray (from Cs-137 and Co-60 sources)and for neutrons (from am-Be source). A reduction of about 10% in the HVL was obtained for the concrete from Damascus in comparison with that from Aleppo, for both neutrons and gammas. One of the other four types of concrete (from Rajo site, mostly Hematite), was found to further reduce the HVL by about 10% for both neutrons and gamma rays.(author)

Isotope effects on nuclear shielding

International Nuclear Information System (INIS)

Hansen, P.E.

1983-01-01

This review concentrates upon empirical trends and practical uses of mostly secondary isotope effects, both of the intrinsic and equilibrium types. The text and the tables are arranged in the following fashion. The most 'popular' isotope effect is treated first, deuterium isotope effects on 13 C nuclear shielding, followed by deuterium on 1 H nuclear shieldings, etc. Focus is thus on the isotopes producing the effect rather than on the nuclei suffering the effect. After a brief treatment of each type of isotope effect, general trends are dealt with. Basic trends of intrinsic isotope effects such as additivity, solvent effects, temperature effects, steric effects, substituent effects and hyperconjugation are discussed. Uses of isotope effects for assignment purposes, in stereochemical studies, in hydrogen bonding and in isotopic tracer studies are dealt with. Kinetic studies, especially of phosphates, are frequently performed by utilizing isotope effects. In addition, equilibrium isotope effects are treated in great detail as these are felt to be new and very important and may lead to new uses of isotope effects. Techniques used to obtain isotope effects are briefly surveyed at the end of the chapter. (author)

Polyolefin-Nanocrystal Composites for Radiation Shielding, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — EIC Laboratories Inc. is proposing a lightweight multifunctional polymer/nanoparticle composite for radiation shielding during long-duration lunar missions. Isolated...

A Statistical Model for Generating a Population of Unclassified Objects and Radiation Signatures Spanning Nuclear Threats

International Nuclear Information System (INIS)

Nelson, K.; Sokkappa, P.

2008-01-01

This report describes an approach for generating a simulated population of plausible nuclear threat radiation signatures spanning a range of variability that could be encountered by radiation detection systems. In this approach, we develop a statistical model for generating random instances of smuggled nuclear material. The model is based on physics principles and bounding cases rather than on intelligence information or actual threat device designs. For this initial stage of work, we focus on random models using fissile material and do not address scenarios using non-fissile materials. The model has several uses. It may be used as a component in a radiation detection system performance simulation to generate threat samples for injection studies. It may also be used to generate a threat population to be used for training classification algorithms. In addition, we intend to use this model to generate an unclassified 'benchmark' threat population that can be openly shared with other organizations, including vendors, for use in radiation detection systems performance studies and algorithm development and evaluation activities. We assume that a quantity of fissile material is being smuggled into the country for final assembly and that shielding may have been placed around the fissile material. In terms of radiation signature, a nuclear weapon is basically a quantity of fissile material surrounded by various layers of shielding. Thus, our model of smuggled material is expected to span the space of potential nuclear weapon signatures as well. For computational efficiency, we use a generic 1-dimensional spherical model consisting of a fissile material core surrounded by various layers of shielding. The shielding layers and their configuration are defined such that the model can represent the potential range of attenuation and scattering that might occur. The materials in each layer and the associated parameters are selected from probability distributions that span the

Determining optical and radiation characteristics of cathode ray tubes' glass to be reused as radiation shielding glass

Science.gov (United States)

Zughbi, A.; Kharita, M. H.; Shehada, A. M.

2017-07-01

A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide.

Gamma radiation shielding and optical properties measurements of zinc bismuth borate glasses

International Nuclear Information System (INIS)

Yasaka, P.; Pattanaboonmee, N.; Kim, H.J.; Limkitjaroenporn, P.; Kaewkhao, J.

2014-01-01

Highlights: • 10ZnO:xBi 2 O 3 :(90−x)B 2 O 3 , (ZBB) glasses were prepared. • Radiation shielding and optical properties were investigated. • Higher 25 mol% of Bi 2 O 3 show better shielding property compared with concretes. • ZBB glasses can develop as a Pb-free radiation shielding material. - Abstract: In this work, the zinc bismuth borate (ZBB) glasses of the composition 10ZnO:xBi 2 O 3 :(90−x)B 2 O 3 (where x = 15, 20, 25 and 30 mol%) were prepared by the melt quenching technique. Their radiation shielding and optical properties were investigated and compared with theoretical calculations. The mass attenuation coefficients of ZBB glasses have been measured at different energies obtained from a Compton scattering technique. The results show a decrease of the mass attenuation coefficient, effective atomic number and effective electron density values with increasing of gamma-ray energies; and good agreements between experimental and theoretical values. The glass samples with Bi 2 O 3 concentrations higher than 25 mol% (25 and 30 mol%) were observed with lower mean free path (MFP) values than all the standard shielding concretes studied. These results are indications that the ZBB glasses in the present study may be developed as a lead-free radiation shielding material in the investigated energy range

Study and installation of concrete shielding in the civil engineering of nuclear construction (1960)

International Nuclear Information System (INIS)

Dubois, F.

1960-01-01

The object of this report is to give technical information about high density concretes which have become very important for radiation biological shielding. The most generally used heavy aggregates (barytes, ilmenite, ferrophosphorus, limonite, magnetite and iron punching) to make these concretes are investigated from the point of view prospecting and physical and chemical characteristics. At first, a general survey of shielding concretes is made involving the study of components, mixing and placing methods, then, a detailed investigation of some high density concretes: barytes concrete, with incorporation of iron punching or iron shot, ferrophosphorus concrete, ilmenite concrete and magnetite concrete, more particularly with regard to grading and mix proportions and testing process. To put this survey in concrete form, two practical designs are described such as they have been carried out at the Saclay Nuclear Station. Specifications are given for diverse concretes and for making the proton-synchrotron 'Saturne' shielding blocks. (author) [fr

Radiation shielding techniques and applications. 3. Analysis of Photon Streaming Through and Around Shield Doors

International Nuclear Information System (INIS)

Barnett, Marvin; Hack, Joe; Nathan, Steve; White, Travis

2001-01-01

Westinghouse Safety Management Solutions (Westinghouse SMS) has been tasked with providing radiological engineering design support for the new Commercial Light Water Reactor Tritium Extraction Facility (CLWR-TEF) being constructed at the Savannah River Site (SRS). The Remote Handling Building (RHB) of the CLWR-TEF will act as the receiving facility for irradiated targets used in the production of tritium for the U.S. Department of Energy (DOE). Because of the high dose rates, approaching 50 000 rads/h (500 Gy/h) from the irradiated target bundles, significant attention has been made to shielding structures within the facility. One aspect of the design that has undergone intense review is the shield doors. The RHB has six shield doors that needed to be studied with respect to photon streaming. Several aspects had to be examined to ensure that the design meets the radiation dose levels. Both the thickness and streaming issues around the door edges were designed and examined. Photon streaming through and around a shield door is a complicated problem, creating a reliance on computer modeling to perform the analyses. The computer code typically used by the Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. The complexity of the geometry within the problem can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles transport through the shield door with transport through the streaming paths, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all important regions of the model are sampled appropriately. The thickness determination used a simple variance reduction technique. In construction, the shield door will not be flush against the wall, so a solid rectangular slab leaves streaming paths around the edges. Administrative controls could be used to control dose to workers; however, 10 CFR 835.1001 states

Determining optical and radiation characteristics of cathode ray tubes' glass to be reused as radiation shielding glass

International Nuclear Information System (INIS)

Zughbi, A.; Kharita, M.H.; Shehada, A.M.

2017-01-01

A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide. - Highlights: • A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented. • The glass from CRTs used as raw materials for radiation shielding glass. • The resulted glass have good optical properties and stability against radiations.

Verification of shielding effect by the water-filled materials for space radiation in the International Space Station using passive dosimeters

Czech Academy of Sciences Publication Activity Database

Kodaira, S.; Tolochek, R. V.; Ambrožová, Iva; Kawashima, H.; Yasuda, N.; Kurano, M.; Kitamura, H.; Uchihori, Y.; Kobayashi, I.; Hakamada, H.; Suzuki, A.; Kartsev, I. S.; Yarmanova, E. N.; Nikolaev, I. V.; Shurshakov, V. A.

2014-01-01

Roč. 53, č. 1 (2014), s. 1-7 ISSN 0273-1177 Institutional support: RVO:61389005 Keywords : space radiation dosimetry * water shield * dose reduction * passive dosimeters * CR-39 * TLD Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.358, year: 2014

Radiation Exposure Analyses Supporting the Development of Solar Particle Event Shielding Technologies

Science.gov (United States)

Walker, Steven A.; Clowdsley, Martha S.; Abston, H. Lee; Simon, Hatthew A.; Gallegos, Adam M.

2013-01-01

NASA has plans for long duration missions beyond low Earth orbit (LEO). Outside of LEO, large solar particle events (SPEs), which occur sporadically, can deliver a very large dose in a short amount of time. The relatively low proton energies make SPE shielding practical, and the possibility of the occurrence of a large event drives the need for SPE shielding for all deep space missions. The Advanced Exploration Systems (AES) RadWorks Storm Shelter Team was charged with developing minimal mass SPE storm shelter concepts for missions beyond LEO. The concepts developed included "wearable" shields, shelters that could be deployed at the onset of an event, and augmentations to the crew quarters. The radiation transport codes, human body models, and vehicle geometry tools contained in the On-Line Tool for the Assessment of Radiation In Space (OLTARIS) were used to evaluate the protection provided by each concept within a realistic space habitat and provide the concept designers with shield thickness requirements. Several different SPE models were utilized to examine the dependence of the shield requirements on the event spectrum. This paper describes the radiation analysis methods and the results of these analyses for several of the shielding concepts.

Evaluation of gamma ray durability and its application of shielded RF tags

International Nuclear Information System (INIS)

Teraura, Nobuyuki; Ito, Kunio; Kobayashi, Daisuke; Sakurai, Kouichi

2015-01-01

In this study, the RF (Radio Frequency) tag with radiation shield is developed and its gamma ray durability is evaluated. RFID (RF Identification) is a radio-wave-based identification technology that can be used for various items. RF tags find use in many applications, including item tracing, access control, etc. RF tags can be classified as active RF tags, which have inbuilt voltaic cells, and passive RF tags without these cells. Passive RF tags, known for their low price and durability, are used in various fields. For instance, they are used for equipment maintenance in factories and thermal power plants. Several frequencies are used for RF tags. Further, RF tagging on the UHF (Ultra High Frequency) frequencies allows a communication range of approximately 10 m, and thus, remote reading is possible. When used in radiation environments such as in nuclear power plants, remote reading can contribute to the reduction of radiation exposure. However, because semiconductors are the primary elements used in the manufacture of RF tags, they can be damaged by radiation, and operational errors can occur. Therefore, this technology has not been used in environments affected by relatively high radiation levels. Therefore, in nuclear power plants, the use of RF tags is limited in areas of low radiation levels. In our study, we develop and manufacture a new RF tag with a radiation shield cover that provides error correction functionality. It is expected that radiation shielded RF tags will improve the radiation-proof feature, and its application range will be expanded. Using the radiation-proof RF tag, we have conducted radiation durability tests. These tests are of two types: one using low energy gamma ray, and the other using high-energy gamma ray. Experimental results are then analyzed. The number of applications for radiation shielded RF tags is considerably increasing, because it can be used in various radiation environments other than nuclear power plants as well, such as

Evaluation of the computer code system RADHEAT-V4 by analysing benchmark problems on radiation shielding

International Nuclear Information System (INIS)

Sakamoto, Yukio; Naito, Yoshitaka

1990-11-01

A computer code system RADHEAT-V4 has been developed for safety evaluation on radiation shielding of nuclear fuel facilities. To evaluate the performance of the code system, 18 benchmark problem were selected and analysed. Evaluated radiations are neutron and gamma-ray. Benchmark problems consist of penetration, streaming and skyshine. The computed results show more accurate than those by the Sn codes ANISN and DOT3.5 or the Monte Carlo code MORSE. Big core memory and many times I/O are, however, required for RADHEAT-V4. (author)

Investigation and assessment of lead slag concrete as nuclear shields

International Nuclear Information System (INIS)

Zaghloul, Y.R.

2009-01-01

The present work is concerned with the efficiency of heavy weight concrete as a shielding material in constructing nuclear installations as well as for radioactive wastes disposal facilities.In this context, lead slag was used as a replacement for fine aggregates in heavy concrete shields that include local heavy weight aggregates (namely; barite and ilmenite) as well as normal concrete includes dolomite and sand as coarse and fine aggregates, as a reference. The effect of different percentages of lead slag was investigated to assess the produced lead slag concrete as a nuclear shielding material. The different properties (physical, mechanical and nuclear) of the produced lead slag concrete were investigated. The results obtained showed that increasing the lead slag percentage improving the investigated properties of the different concrete mixes. In addition, ilmenite concrete with 20% lead slag showed the best results for all the investigated properties.

Radiation shield vest and skirt

International Nuclear Information System (INIS)

Maine, G.J.

1982-01-01

A two-piece garment is described which provides shielding for female workers exposed to radiation. The upper part is a vest, overlapping and secured in the front by adjustable closures. The bottom part is a wraparound skirt, also secured by adjustable closures. The two parts overlap, thus providing continuous protection from shoulder to knee and ensuring that the back part of the body is protected as well as the front

Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

International Nuclear Information System (INIS)

McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.

1980-01-01

The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

International Nuclear Information System (INIS)

Sato, Satoshi

2003-09-01

In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

The extreme condition analyzing for NEMPI shielding of electronic system in high-intensity pulsed radiation diagnosing

International Nuclear Information System (INIS)

Cheng Xiaolei; Liu Fang; Ouyang Xiaoping

2012-01-01

The difficulty for estimating the NEMPI (electromagnetic pulsed interference caused by the nuclear reaction) on the electronic system in high-intensity pulsed radiation diagnosing is analyzed in this article. To solve the difficulty, a method called 'Extreme Condition Analyzing' is presented for estimating the NEMPI conservatively and reliably. Through an extreme condition hypothesizing which could be described as 'Entire Coupling of Electric Field Energy', the E max (maximum electric field intensity which could be endured by the electronic system in the high-intensity pulsed radiation) could be figured out without any other information of the EMP caused by the nuclear reaction. Then a feasibility inspection is introduced, to confirm that the EMPI shielding request according to E max is not too extreme to be achieved. (authors)

Nuclear shielding of openings in ITER Tokamak building

Energy Technology Data Exchange (ETDEWEB)

Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

Angelo, J.P. Jr.; Buden, D.

1982-01-01

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

The evaluation of the radiation shielding ability of lead glass

International Nuclear Information System (INIS)

Tsuda, Keisuke; Fukushi, Masahiro; Myojoyama, Atsushi; Kitamura, Hideaki; Nakaya, Giichiro; Hassan, Nabil; Inoue, Kazumasa; Kimura, Junichi; Sawaguchi, Masato; Kinase, Sakae; Saito, Kimiaki

2008-01-01

Positron emission tomography (PET) scanning with the tracer 2-[F-18] Fluoro-2deoxy-D-glucose (FDG) is widely used in the clinical PET. However, the photon energy used in the PET scans is considerably higher than that of the X-rays traditionally used in the diagnoses. The radiation protection in the PET institution, therefore, is the remaining problem. Meanwhile, lead glass has attracted considerable attention as a radiation-shielding material for the PET institution. The aim of the present study was to evaluate the radiation-shielding ability of the lead glass against the positron emitters. The shielding ability evaluations were done both in the actual experiments and in the Monte Carlo simulation. The lead glass, the object of evaluation in this study, proved to have sufficient protective effect. The development and the spread of a thinner and lighter lead glass with the same effective dose transmission factor should be expected in the near future. (author)

REACTOR SHIELD

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Improved Metal-Polymeric Laminate Radiation Shielding, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — In this proposed Phase II program, builds on the phase I feaibility where a multifunctional lightweight radiation shield composite was developed and fabricated. This...

Nuclear data requirements for fusion reactor shielding

International Nuclear Information System (INIS)

Abdou, M.A.

1979-01-01

The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

TFTR radiation contour and shielding efficiency measurements during D-D operations

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Hwang, D.; Lewis, M.; Levine, J.; Ku, L.P.; Rule, K.; Hajnal, F.

1994-11-01

Extensive neutron and gamma radiation contour, shielding efficiency, and spectral measurements were performed during high power TFTR D-D operations at the tokamak Test Cell inner walls, ceiling, roof, and outer walls, in nearby control rooms, work areas, and personnel pathways, outdoors along the site fence at 125 m, and out to the nearest property lines at 180 m. The results confirmed that the efficiency of the basic radiation shielding was sufficient to allow the TFTR D-T experimental plan, and provide empirical guidance for simulating the radiation fields of future fusion reactors

An Evaluation on Radiation Shielding and Activation Properties of ISOL-bunker Structural Materials for Radiation Safety in RAON Accelerator

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Hyun; Kim, Song Hyun; Woo, Myeong Hyeon; Lee, Jae Yong; Kim, Jong Woo; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Nam, Shin Woo [Institute for Basic Science, Daejeon (Korea, Republic of)

2015-10-15

RAON heavy ion accelerator has been designed by the Institute for Basic Science (IBS). ISOL is one of RAON facilities to generate and separate rare isotopes. For generating rare isotopes, high intensity proton beam, which has 70 MeV energy, is induced into UCx target. From this reaction, lots of neutrons are concomitantly generated. To meet our design goal, it was required that the structural material of ISOL-bunker should be carefully selected. In this study, to select the structural material which has lower activation property with higher performance for radiation shielding, following aspects were evaluated: (i) residual dose, (ii) radioactive wastes, and (iii) shielding performance in ISOL-bunker. In this study, to effectively design the radiation shielding of the RAON ISOL-bunker, two methods were proposed. No.1 strategy is a method to replace the normal concrete to specific concretes. No.2 strategy is to design dual-layer radiation shields that a specific shielding material is located inner side of the normal concrete. Using the strategies, performance evaluations were evaluated for three aspects, which are residual dose, radioactive waste, and prompt radiation. The results show that the residual radiation can be effectively reduced using B{sub 4}C, borated polyethylene and polyethylene with No.2 strategy. Also, the colemanite concrete and B{sub 4}C shielding give a good ability to reduce the radioactive wastes.

A study on the characteristics of modified and novolac type epoxy resin based neutron shielding material

Energy Technology Data Exchange (ETDEWEB)

Cho, Soo Haeng; Hong, Sun Seok; Oh, Seung Chul; Do, Jae Bum [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. In this study, we developed modified and novolac type epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, prolonged time heat resistance, thermal and mechanical properties were evaluated experimently. (author). 31 refs., 27 figs., 16 tabs.

Shielding estimation for nuclear medicine therapy ward: our experience

International Nuclear Information System (INIS)

Skopljak-Beganovic, A.; Kucukalic-Selimovic, E.; Beganovic, A.; Drljevic, A.

2008-01-01

Full text: The aim of this study was to calculate and estimate the shielding thickness for a new Nuclear Medicine Therapy Ward. Parameters available for shielding calculation were: ground plan of the ward, radionuclides planned for use, maximum administered activity of I-131, maximum delivered activity of I-131 to the ward per week, average time spent in the hospital after the treatment. The most hazardous and most commonly used radioisotope is I-131. The target dose that needs to be met for occupationally exposed workers is 0.3 mSv per year. There are several factors that could be changed in order to achieve this value: distance from the source, shielding thickness, angle of incidence, occupational and usage factors. The maximum dose rate at 1 meter from the thyroid gland of the patient was considered to be 100 mSv/h. The distances and incidence angles could not be changed since these vales were predetermined in the ground plan. Different usage and occupational factors were used for different rooms in the ward. We used occupational factor 1 for the bed and 1/6 for the bathroom, and usage factor 1 for nurses' room and patient room and 1/6 for the corridors, etc. The easiest way of calculating dose attenuation in material was by introducing the HVL and TVL for broad beams. TVL and HVL were taken from the graph.The results show that shielding thickness should be in the range of 3 mmPb for room doors to 30 mmPb for the wall adjacent to the nurse's office. Most of the walls are 20 mmPb thick. These values were calculated using conservative assumptions and are more then enough to protect staff, patients and public from external radiation. If the construction cannot support the weight of lead some rearrangements regarding patient positions could be made. (author)

Glasses impregnated with lead for radiation shielding

International Nuclear Information System (INIS)

Abd El Monem, A.M.; Kansouh, W.A.; Megahid, R.M.; Ismail, A.L.; Awad, E.M.

2005-01-01

The attenuation properties of glasses with different concentration of lead have been investigated for the attenuation of gamma-rays from cesium-137 and for total gamma rays using a beam of neutrons and gamma rays emitted from californium-252 source. Measurements have been performed using a gamma-ray spectrometer with Nal(T1) detector for gamma-rays emitted from 137 Cs and a neutron/gamma spectrometer with stilbene scintillator for measurement of total gamma-rays from 252 Cf neutron source. The latter applied the pulse shape discrimination technique to distinguish between recoil proton and recoil electron pulses. The obtained results given the form displayed pulse height spectra and attenuation relations which were used to derive the linear attenuation coefficient (μ), and the mass attenuation coefficient (mu/p) of the investigated glasses. In addition, calculations were performed to determine the attenuation properties of glass shields under investigation using XCOM code given by the others. A comparison of the shielding properties of these glasses with some standard shielding materials indicated that, the investigated glasses process the shielding advantages required for different nuclear technology applications

Efficient Radiation Shielding Through Direct Metal Laser Sintering

Data.gov (United States)

National Aeronautics and Space Administration — We have developed a method for efficient component-level radiation shielding that can be printed by direct metal laser sintering (DMLS) from files generated by the...

Radiation dose reduction at a price: the effectiveness of a thyroid shield during head CT scanning

International Nuclear Information System (INIS)

Fu Qiang; Lu Tao; Zhang Ling

2008-01-01

Objective: To assess radiation dose to the thyroid in patients undergoing head CT scanning and to evaluate dose reduction to the thyroid by load shielding. Methods: A post-morterm was scanned by different model and study was undertaken to evaluate the dose reduction by thyroid lead shields and assess their practicality in a clinical setting. (a)No thyroid shields and (b) thyroid shield. One thermoluminescent dosimeters (TLDs)were placed over the thyroid gland center, A thyroid lead shield (Pb eq 0.5mm)was placed around the neck of post-morterm. Scan parameter, CTDIw and DLP were recorded. Results: (a) 0.207mSv; (b) 0.085mSv. A mean effective radiation dose reduction of 58% was seen in the shielded versus the unshielded. Conclusion: Thyroid exposure to scattered radiation from head CT scanning only once is associated with a low but not negligible risk of cancer, but accumulatived doses to the thyroid are serious, highlighting the need for increased awareness of patient radiation protection. Thyroid lead shielding yields significant radiation protection, which should be used routinely during head CT scan. (authors)

An analysis of exposure dose on hands of radiation workers using a Monte Carlo simulation in nuclear medicine

Energy Technology Data Exchange (ETDEWEB)

Jang, Dong Gun [Dept. of Nuclear Medicine, Dongnam Institute of Radiological and Medical Sciences Cancer Center, Pusan (Korea, Republic of); Kang, SeSik; Kim, Jung Hoon; KIm, Chang Soo [Dept. of Radiological Science, College of Health Sciences, Catholic University, Pusan (Korea, Republic of)

2015-12-15

Workers in nuclear medicine have performed various tasks such as production, distribution, preparation and injection of radioisotope. This process could cause high radiation exposure to workers’ hand. The purpose of this study was to investigate shielding effect for r-rays of 140 and 511 keV by using Monte-Carlo simulation. As a result, it was effective, regardless of lead thickness for radiation shielding in 140 keV r-ray. However, it was effective in shielding material with thickness of more than only 1.1 mm in 511 keV r-ray. And also it doesn’t effective in less than 1.1 mm due to secondary scatter ray and exposure dose was rather increased. Consequently, energy of radionuclide and thickness of shielding materials should be considered to reduce radiation exposure.

Gravity Scaling of a Power Reactor Water Shield

International Nuclear Information System (INIS)

Reid, Robert S.; Pearson, J. Boise

2008-01-01

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

Nuclear magnetic shielding tensors of 207Pb2+ in Pb(NO3)2

International Nuclear Information System (INIS)

Lutz, O.; Nolle, A.

1980-01-01

The NMR signals of 207 Pb were observed in a single crystal of Pb(NO 3 ) 2 and could be assigned to the four different Pb 2+ sites by the dependence of the linewidths on the orientation. Four different nuclear magnetic shielding tensors with equal principal values but with different characteristic vectors could be determined. The symmetry of the shielding tensors is in agreement with the symmetry at the Pb 2+ sites. It is shown, that intermolecular contributions can not account for the anisotropy of the nuclear magnetic shielding, which is 3 0 / 00 of the isotropic absolute magnetic shielding. (orig.)

Graphic system for the analysis of representation of a complex three-dimensional configuration for radiation shield calculation

International Nuclear Information System (INIS)

Berezhkov, A.B.; Gordeeva, E.K.; Mazanov, V.L.; Solov'ev, V.Yu.; Ryabov, A.V.; Khokhlov, V.F.; Shejno, I.N.

1987-01-01

Programs for obtaining phantom images when calculating the radiation shield structure for nuclear-engineering plants, using computer graphics, are developed. Programs are designed to accompany calculational investigations using the SUPER2/RRI3-PICSCH program and ZAMOK-TOMOGRAF program comutering complexes. Design geometry techniques, allowing to present three-dimensional object in the form of two-dimensional perspective projection to the screen plane, are realized in the programs

Effectiveness of Bismuth Shield to Reduce Eye Lens Radiation Dose Using the Photoluminescence Dosimetry in Computed Tomography

International Nuclear Information System (INIS)

Jung, Mi Young; Kweon, Dae Cheol; Kwon, Soo Il

2009-01-01

The purpose of our study was to determine the eye radiation dose when performing routine multi-detector computed tomography (MDCT). We also evaluated dose reduction and the effect on image quality of using a bismuth eye shield when performing head MDCT. Examinations were performed with a 64MDCT scanner. To compare the shielded/unshielded lens dose, the examination was performed with and without bismuth shielding in anthropomorphic phantom. To determine the average lens radiation dose, we imaged an anthropomorphic phantom into which calibrated photoluminescence glass dosimeter (PLD) were placed to measure the dose to lens. The phantom was imaged using the same protocol. Radiation doses to the lens with and without the lens shielding were measured and compared using the Student t test. In the qualitative evaluation of the MDCT scans, all were considered to be of diagnostic quality. We did not see any differences in quality between the shielded and unshielded brain. The mean radiation doses to the eye with the shield and to those without the shield were 21.54 versus 10.46 mGy, respectively. The lens shield enabled a 51.3% decrease in radiation dose to the lens. Bismuth in-plane shielding for routine eye and head MDCT decreased radiation dose to the lens without qualitative changes in image quality. The other radiosensitive superficial organs specifically must be protected with shielding.

Application of gypsum as shielding against low-energy X-radiation in the radiodiagnosis area

International Nuclear Information System (INIS)

Lins, J.A.G.; Lima, F.R.A.; Santos, M.A.P. dos; Oliveira, D.N.S. de; Silva, V.H.F.F. da

2017-01-01

In recent years, materials such as lead, concrete and iron have been studied for use as shielding for ionizing radiations of different energies in radiative installations. In the radiodiagnosis area, lead and barite are the most used materials as shielding. However, for beams of low energy X-radiation, such as in mammography and dentistry, the gypsum material may be used. This study aims to verify the feasibility of the use of gypsum as shielding for low-energy X-ray using standardized dental X-ray beams in a metrology laboratory. The project will allow a better understanding in the study of gypsum used as shielding, certifying its use as a good attenuator for low-energy X-ray

Radiation transmission data for radionuclides and materials relevant to brachytherapy facility shielding.

Science.gov (United States)

Papagiannis, P; Baltas, D; Granero, D; Pérez-Calatayud, J; Gimeno, J; Ballester, F; Venselaar, J L M

2008-11-01

To address the limited availability of radiation shielding data for brachytherapy as well as some disparity in existing data, Monte Carlo simulation was used to generate radiation transmission data for 60Co, 137CS, 198Au, 192Ir 169Yb, 170Tm, 131Cs, 125I, and 103pd photons through concrete, stainless steel, lead, as well as lead glass and baryte concrete. Results accounting for the oblique incidence of radiation to the barrier, spectral variation with barrier thickness, and broad beam conditions in a realistic geometry are compared to corresponding data in the literature in terms of the half value layer (HVL) and tenth value layer (TVL) indices. It is also shown that radiation shielding calculations using HVL or TVL values could overestimate or underestimate the barrier thickness required to achieve a certain reduction in radiation transmission. This questions the use of HVL or TVL indices instead of the actual transmission data. Therefore, a three-parameter model is fitted to results of this work to facilitate accurate and simple radiation shielding calculations.

Symbolic math for computation of radiation shielding

International Nuclear Information System (INIS)

Suman, Vitisha; Datta, D.; Sarkar, P.K.; Kushwaha, H.S.

2010-01-01

Radiation transport calculations for shielding studies in the field of accelerator technology often involve intensive numerical computations. Traditionally, radiation transport equation is solved using finite difference scheme or advanced finite element method with respect to specific initial and boundary conditions suitable for the geometry of the problem. All these computations need CPU intensive computer codes for accurate calculation of scalar and angular fluxes. Computation using symbols of the analytical expression representing the transport equation as objects is an enhanced numerical technique in which the computation is completely algorithm and data oriented. Algorithm on the basis of symbolic math architecture is developed using Symbolic math toolbox of MATLAB software. Present paper describes the symbolic math algorithm and its application as a case study in which shielding calculation of rectangular slab geometry is studied for a line source of specific activity. Study of application of symbolic math in this domain evolves a new paradigm compared to the existing computer code such as DORT. (author)

Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

CERN Document Server

Žukauskaitėa, A; Plukienė, R; Ridikas, D

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

Reactor head shielding apparatus

International Nuclear Information System (INIS)

Schukei, G.E.; Roebelen, G.J.

1992-01-01

This patent describes a nuclear reactor head shielding apparatus for mounting on spaced reactor head lifting members radially inwardly of the head bolts. It comprises a frame of sections for mounting on the lifting members and extending around the top central area of the head, mounting means for so mounting the frame sections, including downwardly projecting members on the frame sections and complementary upwardly open recessed members for fastening to the lifting members for receiving the downwardly projecting members when the frame sections are lowered thereto with lead shielding supported thereby on means for hanging lead shielding on the frame to minimize radiation exposure or personnel working with the head bolts or in the vicinity thereof

Several problems in accelerator shielding study

International Nuclear Information System (INIS)

Nakamura, Takashi; Hirayama, Hideo; Ban, Shuichi.

1980-01-01

Recently, the utilization of accelerators has increased rapidly, and the increase of accelerating energy and beam intensity is also remarkable. The studies on accelerator shielding have become important, because the amount of radiation emitted from accelerators increased, the regulation of the dose of environmental radiation was tightened, and the cost of constructing shielding rose. As the plans of constructing large accelerators have been made successively, the survey on the present state and the problems of the studies on accelerator shielding was carried out. Accelerators are classified into electron accelerators and proton accelerators in view of the studies on shielding. In order to start the studies on accelerator shielding, first, the preparation of the cross section data is indispensable. The cross sections for generating Bremsstrahlung, photonuclear reactions generating neutrons, generation of neutrons by hadrons, nuclear reaction of neutrons and generation of gamma-ray by hadrons are described. The generation of neutrons and gamma-ray as the problems of thick targets is explained. The shielding problems are complex and diversified, but in this paper, the studies on the shielding, by which basic data are obtainable, are taken up, such as beam damping and side wall shielding. As for residual radioactivity, main nuclides and the difference of residual radioactivity according to substances have been studied. (J.P.N.)

Radiation Shielding Systems Using Nanotechnology

Science.gov (United States)

Chen, Bin (Inventor); McKay, Christoper P. (Inventor)

2011-01-01

A system for shielding personnel and/or equipment from radiation particles. In one embodiment, a first substrate is connected to a first array or perpendicularly oriented metal-like fingers, and a second, electrically conducting substrate has an array of carbon nanostructure (CNS) fingers, coated with an electro-active polymer extending toward, but spaced apart from, the first substrate fingers. An electric current and electric charge discharge and dissipation system, connected to the second substrate, receives a current and/or voltage pulse initially generated when the first substrate receives incident radiation. In another embodiment, an array of CNSs is immersed in a first layer of hydrogen-rich polymers and in a second layer of metal-like material. In another embodiment, a one- or two-dimensional assembly of fibers containing CNSs embedded in a metal-like matrix serves as a radiation-protective fabric or body covering.

RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

International Nuclear Information System (INIS)

Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

1989-03-01

A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

Development of Neutron Shielding Material for Cask and Accelerator

International Nuclear Information System (INIS)

Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

2008-01-01

The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

Evaluation of the gamma radiation shielding parameters of bismuth modified quaternary glass system

Science.gov (United States)

Kaur, Parminder; Singh, K. J.; Thakur, Sonika

2018-05-01

Glasses modified with heavy metal oxides (HMO) are an interesting area of research in the field of gamma-ray shielding. Bismuth modified lithium-zinc-borate glasses have been studied whereby bismuth oxide is added from 0 to 50 mol%. The gamma ray shielding properties of the glasses were evaluated at photon energy 662 keV with the help of XMuDat computer program by using the Hubbell and Seltzer database. Various gamma ray shielding parameters such as attenuation coefficient, shield thickness in terms of half and tenth value layer, effective atomic number have been studied in this work. A useful comparison of this glass system has been made with standard radiation shielding concretes viz. ordinary, barite and iron concrete. The glass samples containing 20 to 50 mol% bismuth oxide have shown better gamma ray shielding properties and hence have the potential to become good radiation absorbers.

Radiation shielding analysis of a special linear accelerator for electron beam and X-ray.

Science.gov (United States)

Kang, W G; Pyo, S H; Alkhuraiji, T S; Han, B S; Kang, C M

2017-01-01

The King AbdulAziz City for Science & Technology in the Kingdom of Saudi Arabia plans to build a 10 MeV, 15 kW linear accelerator (LINAC) for electron beam and X-ray. The accelerator will be supplied by EB Tech, Republic of Korea, and the design and construction of the accelerator building will be conducted in the cooperation with EB Tech. This report presents the shielding analysis of the accelerator building using the Monte Carlo N-Particle Transport Code (MCNP). In order to improve the accuracy in estimating deep radiation penetration and to reduce computation time, various variance reduction techniques, including the weight window (WW) method, the deterministic transport (DXTRAN) spheres were considered. Radiation levels were estimated at selected locations in the shielding facility running MCNP6 for particle histories up to 1.0×10+8. The final results indicated that the calculated doses at all selected detector locations met the dose requirement of 50 mSv/yr, which is the United State Nuclear Regulatory Commission (U.S. NRC) requirement.

Dosimetry and shielding

International Nuclear Information System (INIS)

Farinelli, U.

1977-01-01

Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

Growth retardation of paramecium and mouse cells by shielding them from background radiation

International Nuclear Information System (INIS)

Kawanishi, Masanobu; Okuyama, Katsuyuki; Shiraishi, Kazunori; Matsuda, Yatsuka; Taniguchi, Ryoichi; Shiomi, Nobuyuki; Yonezawa, Morio; Yagi, Takashi

2012-01-01

In the 1970s and 1980s, Planel et al. reported that the growth of paramecia was decreased by shielding them from background radiation. In the 1990s, Takizawa et al. found that mouse cells displayed a decreased growth rate under shielded conditions. The purpose of the present study was to confirm that growth is impaired in organisms that have been shielded from background radiation. Radioprotection was produced with a shielding chamber surrounded by a 15 cm thick iron wall and a 10 cm thick paraffin wall that reduced the γ ray and neutron levels in the chamber to 2% and 25% of the background levels, respectively. Although the growth of Paramecium tetraurelia was not impaired by short-term radioprotection (around 10 days), which disagreed with the findings of Planel et al., decreased growth was observed after long-term (40-50 days) radiation shielding. When mouse lymphoma L5178Y cells were incubated inside or outside of the shielding chamber for 7 days, the number of cells present on the 6th and 7th days under the shielding conditions was significantly lower than that present under the non-shielding conditions. These inhibitory effects on cell growth were abrogated by the addition of a 137 Cs γ-ray source disk to the chamber. Furthermore, no growth retardation was observed in XRCC4-deficient mouse M10 cells, which display impaired DNA double strand break repair. (author)

Synchrotron radiation shielding design for the Brockhouse sector at the Canadian light source

International Nuclear Information System (INIS)

Bassey, Bassey; Moreno, Beatriz; Gomez, Ariel; Ahmed, Asm Sabbir; Ullrich, Doug; Chapman, Dean

2014-01-01

At the Canadian Light Source (CLS), the plans for the construction of three beamlines under the Brockhouse Project are underway. The beamlines, to be classified under the CLS Phase III beamlines, will comprise of a wiggler and an undulator, and will be dedicated to x-ray diffraction and scattering experiments. The energy range of these beamlines will be 7–22 keV (low energy wiggler beamline), 20–94 keV (high energy wiggler beamline), and 5–21 keV (undulator beamline). The beamlines will have a total of five hutches. Presented is the shielding design against target scattered white and monochromatic synchrotron radiations for these beamlines. The shielding design is based on: scatter target material-water, dose object-anthropomorphic phantom of the adult human (anteroposterior-AP geometry), and shielding thicknesses of steel and lead that will drop the radiation leakage from the hutches to below 0.5 μSv/h. - Highlights: • The Brockhouse project will add 3 new beamlines at the Canadian Light Source (CLS). • The shielding design against synchrotron radiation was required for these beamlines. • We have completed the required shielding design. • Our design will reduce radiation leakage to <0.5 μSv/h; CLS requires 1.0 μSv/h

Prevalence of Protective Shielding Utilization for Radiation Dose Reduction in Adult Patients Undergoing Body Scanning Using Computed Tomography.

Science.gov (United States)

Safiullah, Shoaib; Patel, Roshan; Uribe, Brittany; Spradling, Kyle; Lall, Chandana; Zhang, Lishi; Okhunov, Zhamshid; Clayman, Ralph V; Landman, Jaime

2017-10-01

Ionizing radiation is implicated in nearly 2% of malignancies in the United States; radiation shields prevent unnecessary radiation exposure during medical imaging. Contemporary radiation shield utilization for adult patients in the United States is poorly defined. Therefore, we evaluated the prevalence of protective shielding utilization in adult patients undergoing CT scans in United States' hospitals. An online survey was sent to established radiology departments randomly selected from the 2015 American Hospital Association Guide. Radiology departments conducting adult CT imaging were eligible; among 370 eligible departments, 215 departments accepted the study participation request. Questions focused on shielding practices during CT imaging of the eyes, thyroid, breasts, and gonads. Prevalence data were stratified per hospital location, size, and type. Main outcomes included overall protective shielding utilization, respondents' belief and knowledge regarding radiation safety, and organ-specific shielding prevalence. Sixty-seven of 215 (31%) hospitals completed the survey; 66 (99%) reported familiarity with the ALARA (as low as reasonably achievable) principle and 56 (84%) affirmed their belief that shielding is beneficial. Only 60% of hospitals employed shielding during CT imaging; among these institutions, shielding varied based on CT study: abdominopelvic CT (13, 33%), head CT (33, 83%), or chest CT (30, 75%). Among surveyed hospitals, 40% do not utilize CT shielding despite the majority acknowledging the ALARA principle and agreeing that shielding is a beneficial practice. Failure to address the low prevalence of protective shielding may lead to poor community health due to increased risk of radiation-related cancers.

Problems in radiation shielding calculations with Monte Carlo methods

International Nuclear Information System (INIS)

Ueki, Kohtaro

1985-01-01

The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)

The optimum shielding for a power reactor using local components

International Nuclear Information System (INIS)

AlHajali, S.; Kharita, M. H.; Yousef, S.; Naoom, B.; Al-Nassar, M.

2009-07-01

Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

Monte Carlo simulations for the space radiation superconducting shield project (SR2S).

Science.gov (United States)

Vuolo, M; Giraudo, M; Musenich, R; Calvelli, V; Ambroglini, F; Burger, W J; Battiston, R

2016-02-01

Astronauts on deep-space long-duration missions will be exposed for long time to galactic cosmic rays (GCR) and Solar Particle Events (SPE). The exposure to space radiation could lead to both acute and late effects in the crew members and well defined countermeasures do not exist nowadays. The simplest solution given by optimized passive shielding is not able to reduce the dose deposited by GCRs below the actual dose limits, therefore other solutions, such as active shielding employing superconducting magnetic fields, are under study. In the framework of the EU FP7 SR2S Project - Space Radiation Superconducting Shield--a toroidal magnetic system based on MgB2 superconductors has been analyzed through detailed Monte Carlo simulations using Geant4 interface GRAS. Spacecraft and magnets were modeled together with a simplified mechanical structure supporting the coils. Radiation transport through magnetic fields and materials was simulated for a deep-space mission scenario, considering for the first time the effect of secondary particles produced in the passage of space radiation through the active shielding and spacecraft structures. When modeling the structures supporting the active shielding systems and the habitat, the radiation protection efficiency of the magnetic field is severely decreasing compared to the one reported in previous studies, when only the magnetic field was modeled around the crew. This is due to the large production of secondary radiation taking place in the material surrounding the habitat. Copyright © 2016 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

The manufacturing of depleted uranium biological shield components

International Nuclear Information System (INIS)

Metelkin, J.A.

1998-01-01

The unique combination of the physical and mechanical properties of uranium made it possible to manufacture biological shield components of transport package container (TPC) for transportation nuclear power plant irradiated fuel and radionuclides of radiation diagnostic instruments. Protective properties are substantially dependent on the nature radionuclide composition of uranium, that why I recommended depleted uranium after radiation chemical processing. Depleted uranium biological shield (DUBS) has improved specific mass-size characteristics compared to a shield made of lead, steel or tungsten. Technological achievements in uranium casting and machining made it possible to manufacture DUBS components of TPC up to 3 tons of mass and up to 2 metres of the maximum size. (authors)

Evaluation of rubber composites as shielding materials against ionizing radiation

International Nuclear Information System (INIS)

Atia, M.K.

2010-01-01

Styrene-butadiene rubber/lead oxide composites were prepared as γ-radiation shields.The composites were prepared with different concentration of red lead oxide (Pb 3 O 4 ) .The assessment of the linear attenuation coefficient of the SBR/lead oxide composites for γ -rays from 137 Cs 137 γ-radiation point source was studied . The factors affecting the mechanical properties and shielding capacity of the composites were also studied. These factors include the lead oxide concentration, the type of monomers added and the irradiation dose. The styrene-butadiene rubber/lead oxide composites can attain up to about 43% of the shielding capacity of pure lead. The incorporation of high concentrations of lead oxide and the effect of accumulative irradiation doses up to 3000 kGy on the physico-mechanical properties of the composites were studied . These led to hardening of the SBR rubber/lead oxide composites.

Optimisation of the radiation shielding of medical cyclotrons using a genetic algorithm

International Nuclear Information System (INIS)

Mukherjee, Bhaskar

2000-01-01

Effective radiation shielding is imperative for safe operation of modern Medical Cyclotrons producing large activities of short-lived radioisotopes on a commercial basis. The optimal cyclotron shielding design demands a careful balance between the radiological, economical and often the sociopolitical factors. One is required to optimize the cost of radiation protection and the cost of radiological-health detriment. The cost of radiation protection depends explicitly on a) the nature of the radiation field produced by the cyclotron, b) the cyclotron operation condition, c) the cost of shielding material, d) the level of dose reduction, e) the projected net revenue from the sale of the radioisotopes, and f) the depreciation rate of the cyclotron facility. The Genetic Algorithm (GA) is used for a cost -benefit analysis of this problem. The GA is a mathematical technique that emulates the Darwinian Evolution paradigm. It is ideally suited to search for a global optimum in a large multi-dimensional solution space, having demonstrated strength compared to the classical analytical methods. Furthermore the GA method runs on a PC in a Windows environment. This paper highlights an interactive spreadsheet macro program for the cost benefit analysis of the optimize Medical Cyclotron shielding using a GA search engine. (author)

Radiation and shielding around beam absorbers

International Nuclear Information System (INIS)

Hurkmans, A.; Maas, R.

1978-12-01

During operational conditions it is anticipated that a fair amount of the total available beam power is dumped in either the slit system on one of the beam dumps. Thses beam absorbers therefore become strong radioactive sources. The radiation level due to the absorption of a 100 kW electron beam is estimated and the problem of residual activity is treated. Proposed shielding materials are discussed. (C.F.)

Reduction of scatter radiation during transradial percutaneous coronary angiography: a randomized trial using a lead-free radiation shield.

Science.gov (United States)

Politi, Luigi; Biondi-Zoccai, Giuseppe; Nocetti, Luca; Costi, Tiziana; Monopoli, Daniel; Rossi, Rosario; Sgura, Fabio; Modena, Maria Grazia; Sangiorgi, Giuseppe M

2012-01-01

Occupational radiation exposure is a growing problem due to the increasing number and complexity of interventional procedures performed. Radial artery access has reduced the number of complications at the price of longer procedure duration. Radpad® scatter protection is a sterile, disposable bismuth-barium radiation shield drape that should be able to decrease the dose of operator radiation during diagnostic and interventional procedures. Such radiation shield has never been tested in a randomized study in humans. Sixty consecutive patients undergoing coronary angiography by radial approach were randomized 1:1 to Radpad use versus no radiation shield protection. The sterile shield was placed around the area of right radial artery sheath insertion and extended medially to the patient trunk. All diagnostic procedures were performed by the same operator to reduce variability in radiation absorption. Radiation exposure was measured blindly using thermoluminescence dosimeters positioned at the operator's chest, left eye, left wrist, and thyroid. Despite similar fluoroscopy time (3.52 ± 2.71 min vs. 3.46 ± 2.77 min, P = 0.898) and total examination dose (50.5 ± 30.7 vs. 45.8 ± 18.0 Gycm(2), P = 0.231), the mean total radiation exposure to the operator was significantly lower when Radpad was utilized (282.8 ± 32.55 μSv vs. 367.8 ± 105.4 μSv, P Radpad utilization at all body locations ranging from 13 to 34% reduction. This first-in-men randomized trial demonstrates that Radpad significantly reduces occupational radiation exposure during coronary angiography performed through right radial artery access. Copyright © 2011 Wiley Periodicals, Inc.

Space Station Validation of Advanced Radiation-Shielding Polymeric Materials, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — In Subtopic X11.01, NASA has identified the need to develop advanced radiation-shielding materials and systems to protect humans from the hazards of space radiation...

Investigation of ionizing radiation shielding effectiveness of decorative building materials used in Bangladeshi dwellings

International Nuclear Information System (INIS)

Yesmin, Sabina; Sonker Barua, Bijoy; Uddin Khandaker, Mayeen; Tareque Chowdhury, Mohammed; Kamal, Masud; Rashid, M.A.; Miah, M.M.H.; Bradley, D.A.

2017-01-01

Following the rapid growing per capita income, a major portion of Bangladeshi dwellers is upgrading their non-brick houses by rod-cement-concrete materials and simultaneously curious to decorate the houses using luxurious marble stones. Present study was undertaken to investigate the gamma-ray attenuation co-efficient of decorative marble materials leading to their suitability as shielding of ionizing radiation. A number of commercial grades decorative marble stones were collected from home and abroad following their large-scale uses. A well-shielded HPGe γ-ray spectrometer combined with associated electronics was used to evaluate the mass attenuation coefficients of the studied materials for high energy photons. Some allied parameters such as half-value layer and radiation protection efficacy of the investigated marbles were calculated. The results showed that among the studied samples, the marble ‘Carrara’ imported from Italy is suitable to be used as radiation shielding material. - Highlights: • Studies of decorative building materials for shielding of ionizing radiation. • High energy photon beam were used to obtain various interaction properties. • Marble stone ‘Carrara’ from Italy shows suitability to be used as shielding material.

An innovative method for on-power radiometry of end-shields of nuclear power plants

International Nuclear Information System (INIS)

Kumar, Gaurav; Gupta, Pankaj; Nawal, Shriram; Gautam, Mahesh; Kakkar, Aman Deep; Yadav, Umed

2012-01-01

Every lndian PHWR reactor calandria is sandwiched within a pair of shield on either side. These shields are perpendicular to the coaxial axis of calandria and are called end-shields. These provide shielding from leakage radiation from reactor core in escaping out to Fuelling Machine vault, thereby significantly reducing the dose rates in the vaults. This has got a direct impact on radiation field in accessible areas. By maintaining low dose rates in accessible areas, the individual and collective doses of radiation workers can be effectively controlled well within the stipulated limits. Thus, it is of utmost importance to ensure adequacy of shielding provided by end-shields. In this context, a limited radiometry exercise is executed after filling of end-shields with steel balls and prior to their installation at designated place. This exercise provides limited inputs along the periphery of end-shield due to limited strength of radiation source, its handling provisions and dose constraints to the individual. In order to ascertain an in-depth analysis of shielding adequacy on-power, different methodologies have been adopted and have certain limitation in precisely pinpointing the affected area/location besides limitation on number of locations that can be monitored at a single stretch. To overcome these important anomalies, a computer based setup has been indigenously designed. The setup essentially comprises of a radiation monitor with wide energy, measuring, temperature and humidity range; a custom designed 25 m long compatible cable with suitable connectors; a laptop with additional cooling arrangement; a configurable interfacing software; thermal shielding for the detector and tying/fixing provisions. The radiation monitor after being properly shielded for thermal impacts is installed on the head of Fuelling Machine. It is connected through long cable to a laptop kept at Fuelling Machine service area with due cooling provisions (as temperature in the area will

Composites with carbon nanotube for radiation shielding application

International Nuclear Information System (INIS)

Fontainha, Críssia C.P.; Nunes, Modesto; Rosas, Víctor A.

2017-01-01

Polymeric composites filled with attenuating metals and functionalized with carbon nanotubes (NTC) are being largely developed. New attenuators materials have been widely investigated for radiation shielding to apply in procedures as interventional radiology, Computed Tomography (CT) and nuclear medicine. In this work composites for radiation attenuation in radiodiagnostic imaging procedures made of inorganic material as filler, by a sol-gel method, in poly(vinylidene fluoride-tryfluorethylene) [P(VDF-TrFE] copolymers that are used as the polymeric matrix. Two different metal attenuators were used as fillers: zirconia stabilized by yttria (8% wt.) and bismuth oxide. Carbon nanotubes were added with different concentrations at the solution of attenuator metal under controlled magnetic stirring. Characterization of composites by FTIR, UV-Vis, DSC and SEM-EDS were carried out. In a previous analysis of radiation attenuation, was used an incident monochromatic X-ray beam from the RIGAKU diffractometer. In this setup, one reference measure is directly exposed to the x-rays being diffracted by single crystal of Si (111). Another measure the attenuated beam is performed with the composite sample under detector. The functionalization of the carbon nanotube of multiple walls (MWNCT) in the in the P(VDF-TrFE) was evaluated. The samples present a good dispersion of the attenuator metal into presence at methacrylic acid. The cheap tube presented better dispersion in the polymer matrix than the 3100 nanotubes. Bismuth oxidation composites showed a better attenuation factor compared to Zirconia stabilized by yttria composites. (author)

Composites with carbon nanotube for radiation shielding application

Energy Technology Data Exchange (ETDEWEB)

Fontainha, Críssia C.P.; Nunes, Modesto; Rosas, Víctor A., E-mail: crissia@gmail.com [Universidade Federal de Minas Gerais (IMA/UFMG), Belo Horizonte, MG (Brazil). Dept. de Anatomia e Imagem; Santos, Adelina P.; Furtado, Clascídia A.; Faria, Luiz O., E-mail: farialo@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Polymeric composites filled with attenuating metals and functionalized with carbon nanotubes (NTC) are being largely developed. New attenuators materials have been widely investigated for radiation shielding to apply in procedures as interventional radiology, Computed Tomography (CT) and nuclear medicine. In this work composites for radiation attenuation in radiodiagnostic imaging procedures made of inorganic material as filler, by a sol-gel method, in poly(vinylidene fluoride-tryfluorethylene) [P(VDF-TrFE] copolymers that are used as the polymeric matrix. Two different metal attenuators were used as fillers: zirconia stabilized by yttria (8% wt.) and bismuth oxide. Carbon nanotubes were added with different concentrations at the solution of attenuator metal under controlled magnetic stirring. Characterization of composites by FTIR, UV-Vis, DSC and SEM-EDS were carried out. In a previous analysis of radiation attenuation, was used an incident monochromatic X-ray beam from the RIGAKU diffractometer. In this setup, one reference measure is directly exposed to the x-rays being diffracted by single crystal of Si (111). Another measure the attenuated beam is performed with the composite sample under detector. The functionalization of the carbon nanotube of multiple walls (MWNCT) in the in the P(VDF-TrFE) was evaluated. The samples present a good dispersion of the attenuator metal into presence at methacrylic acid. The cheap tube presented better dispersion in the polymer matrix than the 3100 nanotubes. Bismuth oxidation composites showed a better attenuation factor compared to Zirconia stabilized by yttria composites. (author)

Radiation-resistant composite for biological shield of personnel

Science.gov (United States)

Barabash, D. E.; Barabash, A. D.; Potapov, Yu B.; Panfilov, D. V.; Perekalskiy, O. E.

2017-10-01

This article presents the results of theoretical and practical justification for the use of polymer concrete based on nonisocyanate polyurethanes in biological shield structures. We have identified the impact of ratio: polymer - radiation-resistant filling compound on the durability and protection properties of polymer concrete. The article expounds regression dependence of the change of basic properties of the aforementioned polymer concrete on the absorbed radiation dose rate. Synergy effect in attenuation of radioactivity release in case of conjoint use of hydrogenous polymer base and radiation-resistant powder is also addressed herein.

Radiation leakage in nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Ando, Yoshio; Miyasaka, Shun-ichi; Takeuchi, Kiyoshi.

1975-01-01

Associated with the radiation leakage in MUTSU occurred in September 1974, this report reviews the shielding design for MUTSU, radiation measurement and inspection activities by a survey group, and 2 dimensional analysis on the behavior of fast neutrons to shielding based on Ssub(N) codes. In the first part, the purpose and the structure of the primary and the secondary shields of MUTSU are briefly illustrated. In the second part, the progress of the series of affairs is explained, starting from zero power criticality experiment, through discovery of radiation leakage in output-increasing test, sending of a survey group for various measurement and inspection, and finally to the conclusion drawn by the survey group. In the third part, various numerical analyses performed to investigate into the leakage are illustrated with their results. The transport codes used were ANISN, TWOTRAN, SPAN, and PALLAS-2DCY. As a result of those inspection and calculation, it was found that the radiation leakage was due to fast neutrons coming through the gap between the reactor vessel and the primary shield. (Aoki, K.)

Reactor vessel head permanent shield

International Nuclear Information System (INIS)

Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

1989-01-01

A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

Evaluation of additional lead shielding in protecting the physician from radiation during cardiac interventional procedures

International Nuclear Information System (INIS)

Chida, Koichi; Zuguchi, Masayuki; Morishima, Yoshiaki; Katahira, Yoshiaki; Chiba, Hiroo

2005-01-01

Since cardiac interventional procedures deliver high doses of radiation to the physician, radiation protection for the physician in cardiac catheterization laboratories is very important. One of the most important means of protecting the physician from scatter radiation is to use additional lead shielding devices, such as tableside lead drapes and ceiling-mounted lead acrylic protection. During cardiac interventional procedures (cardiac IVR), however, it is not clear how much lead shielding reduces the physician dose. This study compared the physician dose [effective dose equivalent (EDE) and dose equivalent (DE)] with and without additional shielding during cardiac IVR. Fluoroscopy scatter radiation was measured using a human phantom, with an ionization chamber survey meter, with and without additional shielding. With the additional shielding, fluoroscopy scatter radiation measured with the human phantom was reduced by up to 98%, as compared with that without. The mean EDE (whole body, mean±SD) dose to the operator, determined using a Luxel badge, was 2.55±1.65 and 4.65±1.21 mSv/year with and without the additional shielding, respectively (p=0.086). Similarly, the mean DE (lens of the eye) to the operator was 15.0±9.3 and 25.73±5.28 mSv/year, respectively (p=0.092). In conclusion, although tableside drapes and lead acrylic shields suspended from the ceiling provided extra protection to the physician during cardiac IVR, the reduction in the estimated physician dose (EDE and DE) during cardiac catheterization with additional shielding was lower than we expected. Therefore, there is a need to develop more ergonomically useful protection devices for cardiac IVR. (author)

Dosimetry and Shielding of X and Gamma Radiation

International Nuclear Information System (INIS)

Oncescu, M.; Panaitescu, I.

1992-01-01

This book covers the following problems: 1. X and Gamma radiations, 2. Interaction of X-ray and gamma radiations with matter, 3. Interaction of electrons with matter, 4. Principles and basic concepts of dosimetry, 5. Ionization dosimetry, 6. Calorimetric chemical and photographic dosimetry, 7. Solid state dosimetry, 8. Computation of dosimetric quantities, 9. Dosimetry in radiation protection, 10. Shielding of X and gamma radiations. The authors, well-known Romanian experts in Radiation Physics and Engineering, gave an up-dated, complete and readable account of this subject matter. The analyses of physical principles and concepts, of materials and instruments and of computational methods and applications are all well balanced to meat the needs of a broad readership

Reconfigurable Patch Antenna Radiations Using Plasma Faraday Shield Effect

OpenAIRE

Barro , Oumar Alassane; Himdi , Mohamed; Lafond , Olivier

2016-01-01

International audience; This letter presents a new reconfigurable antenna associated with a plasma Faraday shield effect. The Faraday shield effect is realized by using a fluorescent lamp. A patch antenna operating at 2.45 GHz is placed inside the lamp. The performance of the reconfigurable system is observed in terms of S11, gain and radiation patterns by simulation and measurement. It is shown that by switching ON the fluorescent lamp, the gain of the antenna decreases and the antenna syste...

Nuclear data needs for fusion reactors

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

A perturbation technique for shield weight minimization

International Nuclear Information System (INIS)

Watkins, E.F.; Greenspan, E.

1993-01-01

The radiation shield optimization code SWAN (Ref. 1) was originally developed for minimizing the thickness of a shield that will meet a given dose (or another) constraint or for extremizing a performance parameter of interest (e.g., maximizing energy multiplication or minimizing dose) while maintaining the shield volume constraint. The SWAN optimization process proved to be highly effective (e.g., see Refs. 2, 3, and 4). The purpose of this work is to investigate the applicability of the SWAN methodology to problems in which the weight rather than the volume is the relevant shield characteristic. Such problems are encountered in shield design for space nuclear power systems. The investigation is carried out using SWAN with the coupled neutron-photon cross-section library FLUNG (Ref. 5)

Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

CERN Document Server

Žukauskaite, A; Plukiene, R; Plukis, A

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

Shielding NSLS-II light source: Importance of geometry for calculating radiation levels from beam losses [Shielding Synchrotron Light Sources: Importance of geometry for calculating radiation levels from beam losses

International Nuclear Information System (INIS)

Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.; Wahl, W.

2016-01-01

Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produces significantly higher neutron component dose to the experimental floor than a lower energy beam injection and ramped operations. Minimizing this dose will require adequate knowledge of where the miss-steered beam can occur and sufficient EM shielding close to the loss point, in order to attenuate the energy of the particles in the EM shower below the neutron production threshold (<10 MeV), which will spread the incident energy on the bulk shield walls and thereby the dose penetrating the shield walls. Designing supplemental shielding near the loss point using the analytic shielding model is shown to be inadequate because of its lack of geometry specification for the EM shower process. To predict the dose rates outside the tunnel requires detailed description of the geometry and materials that the beam losses will encounter inside the tunnel. Modern radiation shielding Monte-Carlo codes, like FLUKA, can handle this geometric description of the radiation transport process in sufficient detail, allowing accurate predictions of the dose rates expected and the ability to show weaknesses in the design before a high radiation incident occurs. The effort required to adequately define the accelerator geometry for these codes has been greatly reduced with the implementation of the graphical interface of FLAIR to FLUKA. This made the effective shielding process for NSLS-II quite accurate and reliable. Lastly, the principles used to provide

Optimal selection for shielding materials by fuzzy linear programming

International Nuclear Information System (INIS)

Kanai, Y.; Miura, N.; Sugasawa, S.

1996-01-01

An application of fuzzy linear programming methods to optimization of a radiation shield is presented. The main purpose of the present study is the choice of materials and the search of the ratio of mixture-component as the first stage of the methodology on optimum shielding design according to individual requirements of nuclear reactor, reprocessing facility, shipping cask installing spent fuel, ect. The characteristic values for the shield optimization may be considered their cost, spatial space, weight and some shielding qualities such as activation rate and total dose rate for neutron and gamma ray (includes secondary gamma ray). This new approach can reduce huge combination calculations for conventional two-valued logic approaches to representative single shielding calculation by group-wised optimization parameters determined in advance. Using the fuzzy linear programming method, possibilities for reducing radiation effects attainable in optimal compositions hydrated, lead- and boron-contained materials are investigated

Experiment and analysis of CASTOR type model cask for verification of radiation shielding

Energy Technology Data Exchange (ETDEWEB)

Hattori, Seiichi; Ueki, Kohtaro.

1988-08-01

The radiation shielding system of CASTOR type cask is composed of the graphite cast iron and the polyethylene lod. The former fomes the cylndrical body of the cask to shield gamma rays and the latter is embeded in the body to shield neutrons. Characteristic of radiation shielding of CASTOR type cask is that zigzag arrangement of the polyethylene lod is adopted to unify the penetrating dose rate. It is necessary to use the three-dimensional analysis code to analyse the shielding performance of the cask with the complicated shielding system precisely. However, it takes too much time as well as too much cost. Therefore, the two-dimensional analysis is usually applied, in which the three-dimensional model is equivalently transformed into the two-dimensional calculation. The reseach study was conducted to verify the application of the two-dimensional analysis, in which the experiment and the analysis using CASTOR type model cask was perfomed. The model cask was manufactured by GNS campany in West Germany and the shielding ability test facilities in CRIEPI were used. It was judged from the study that the two-dimensional analysis is useful means for the practical use.

Post Three Mile Island shielding review - a case history

International Nuclear Information System (INIS)

Isakari, H.H.; Shaw, H.C.

1983-01-01

The radiation shielding review of the Diablo Canyon Nuclear Power Plant was performed in accordance with the requirement of the Three Mile Island Action Plan. The review covered plant shielding and environmental qualification of equipment for spaces and systems which may be used in post-accident operations. Radiation doses during postulated loss-of-coolant accident and high-energy-line-break accident were calculated for equipment located both inside and outside the containment. Vital areas, those requiring post-accident access and occupancy, were identified and their associated dose rates and integrated doses were calculated. It was found that all four of the vital areas (Control Room, Technical Support Center, Switchgear Room, and Emergency Sampling Compartment) are shielded from external sources of radiation sufficiently to permit personnel access and occupancy that would not be unduly limited by the radiation environment caused by the postulated accidents. (author)

Coronary calcium scoring with MDCT: The radiation dose to the breast and the effectiveness of bismuth breast shield

International Nuclear Information System (INIS)

Yilmaz, Mehmet Halit; Yasar, Dogan; Albayram, Sait; Adaletli, Ibrahim; Ozer, Harun; Ozbayrak, Mustafa; Mihmanli, Ismail; Akman, Canan

2007-01-01

Objective: The purpose of our study was to determine the breast radiation dose during coronary calcium scoring with multidetector computerized tomography (MDCT). We also evaluated the degree of dose reduction by using a bismuth breast shield when performing coronary calcium scoring with MDCT. Materials and methods: The dose reduction achievable by shielding the adult (35 years or older) female breasts was studied in 25 women who underwent coronary calcium scoring with MDCT. All examinations were performed with a 16-MDCT scanner. To compare the shielded versus unshielded breast dose, the examinations were performed with (right breast) and without (left breast) breast shielding in all patients. With this technique the superficial breast doses were calculated. To determine the average glandular breast radiation dose, we imaged an anthropomorphic dosimetric phantom into which calibrated dosimeters were placed to measure the dose to the breast. The phantom was imaged using the same protocol. Radiation doses to the breasts with and without the breast shielding were measured and compared using the Student's t-test. Results: The mean radiation doses with and without the breast shield were 5.71 ± 1.1 mGy versus 9.08 ± 1.5 mGy, respectively. The breast shield provided a 37.12% decrease in radiation dose to the breast with shielding. The difference between the dose received by the breasts with and without bismuth shielding was significant, with a p-value of less than 0.001. Conclusion: The high radiation during MDCT greatly exceeds the recommended doses and should not be underestimated. Bismuth in plane shielding for coronary calcium scoring with MDCT decreased the radiation dose to the breast. We recommend routine use of breast shields in female patients undergoing calcium scoring with MDCT

A comparison of dose savings of lead and lightweight aprons for shielding of 99m-Technetium radiation

International Nuclear Information System (INIS)

Warren-Forward, H.; Cardew, P.; Smith, B.; Clack, L.; McWhirter, K.; Johnson, S.; Wessel, K.

2007-01-01

Nuclear medicine technologists (NMTs) have the highest effective doses of radiation among medical workers. With increase in the use of lightweight materials in diagnostic radiography, the aim was to compare the effectiveness of lead and lightweight aprons in shielding from 99m-Technetium ( 99m Tc) gamma rays. The doses received from a scattering phantom to the entrance, 9 cm depth and exit of a phantom were measured with LiF:Mg, Cu, P thermoluminescent dosemeters (TLDs). Doses and spectra were assessed without no shielding, with 0.5-mm lead and lightweight aprons. The lead and lightweight aprons decreased entrance surface doses by 76 and 59%, respectively. The spectral analysis showed that the lightweight apron provided better dose reduction at energies 99m Tc labelled radiopharmaceutical. (authors)

Shielding design for positron emission tomography facility

International Nuclear Information System (INIS)

Abdallah, I.I.

2007-01-01

With the recent advent of readily available tracer isotopes, there has been marked increase in the number of hospital-based and free-standing positron emission tomography (PET) clinics. PET facilities employ relatively large activities of high-energy photon emitting isotopes, which can be dangerous to the health of humans and animals. This coupled with the current dose limits for radiation worker and members of the public can result in shielding requirements. This research contributes to the calculation of the appropriate shielding to keep the level of radiation within an acceptable recommended limit. Two different methods were used including measurements made at selected points of an operating PET facility and computer simulations by using Monte Carlo Transport Code. The measurements mainly concerned the radiation exposure at different points around facility using the survey meter detectors and Thermoluminescent Dosimeters (TLD). Then the set of manual calculation procedures were used to estimate the shielding requirements for a newly built PEF facility. The results from the measurement and the computer simulation were compared to the results obtained from the set manual calculation procedure. In general, the estimated weekly dose at the points of interest is lower than the regulatory limits for the little company of Mary Hospital. Furthermore, the density and the HVL for normal strength concrete and clay bricks are almost similar. In conclusion, PET facilities present somewhat different design requirements and are more likely to require additional radiation shielding. Therefore, existing shields at the little Company of Mary Hospital are in general found to be adequate and satisfactory and additional shielding was found necessary at the new PET facility in the department of Nuclear Medicine of the Dr. George Mukhari Hospital. By use of appropriate design, by implying specific shielding requirements and by maintaining good operating practices, radiation doses to

Study of External Radiation Expose Dose on Hands of Nuclear Medicine Workers

International Nuclear Information System (INIS)

Park, Jun Chul; Pyo, Sung Jae

2012-01-01

The aims of this study are to assess external radiation exposed doses of body and hands of nuclear medicine workers who handle radiation sources, and to measure radiation exposed doses of the hands induced by a whole body bone scan with high frequency and handling a radioactive sources like 99m Tc-HDP and 18 F-FDG in the PET/CT examination. Skillful workers, who directly dispense and inject from radiation sources, were asked to wear a TLD on the chest and ring finger. Then, radiation exposed dose and duration exposed from daily radiation sources for each section were measured by using a pocket dosimeter for the accumulated external doses and the absorbed dose to the hands. In the survey of four medical institutions in Incheon Metropolitan City, only one of four institutions has a radiation dosimeter for local area like hands. Most of institutions uses radiation shielding devices for the purpose of protecting the body trunk, not local area. Even some institutions were revealed not to use such a shielding device. The exposed doses on the hands of nuclear medicine workers who directly handles radioactive sources were approximately twice as much as those on the body. The radiation exposure level for each section of the whole body bone scan with high frequency and that of the PET/CT examination showed that radiation doses were revealed in decreasing order of synthesis of radioactive medicine and installation to a dispensing container, dispensing, administering and transferring. Furthermore, there were statistically significant differences of radiation exposure doses of the hands before and after wearing a syringe shielder in administration of a radioactive sources. In this study, although it did not reach the permissible effective dose for nuclear medicine, the occupational workers were exposed by relatively higher dose level than the non-occupational workers. Therefore, the workers, who closely exposed to radioactive sources should be in compliance with safety

Analysis on the steady-state coherent synchrotron radiation with strong shielding

International Nuclear Information System (INIS)

Li, R.; Bohn, C.L.; Bisognano, J.J.

1997-01-01

There are several papers concerning shielding of coherent synchrotron radiation (CSR) emitted by a Gaussian line charge on a circular orbit centered between two parallel conducting plates. Previous asymptotic analyses in the frequency domain show that shielded steady-state CSR mainly arises from harmonics in the bunch frequency exceeding the threshold harmonic for satisfying the boundary conditions at the plates. In this paper the authors extend the frequency-domain analysis into the regime of strong shielding, in which the threshold harmonic exceeds the characteristic frequency of the bunch. The result is then compared to the shielded steady-state CSR power obtained using image charges

Attenuation characteristics of materials used in radiation protection as radiation shielding

International Nuclear Information System (INIS)

Almeida Junior, Airton T.; Araujo, F.G.S.; Nogueira, M.S.; Santos, M.A.P.

2013-01-01

Crystal glass has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass, in order to minimize exposure to individuals. In this work, ten plates of crystal glass, with dimensions of 20cm x 20cm and range of thicknesses from 0.5 to 2.0 cm, and barite concrete were irradiated with potential constants of 60kV, 80kV, 110kV, 150kV and gamma radiation of 60 Co. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/mA.min) at 1 meter as a function of thickness.Crystal glass has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass, in order to minimize exposure to individuals. In this work, ten plates of crystal glass, with dimensions of 20cm x 20cm and range of thicknesses from 0.5 to 2.0 cm, and barite concrete were irradiated with potential constants of 60kV, 80kV, 110kV, 150kV and gamma radiation of 60 Co. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/mA.min) at 1 meter as a function of thickness. (author)

CHESS upgrade 1995: Improved radiation shielding

International Nuclear Information System (INIS)

Finkelstein, K.

1996-01-01

The Cornell Electron Storage Ring (CESR) stores electrons and positrons at 5.3 GeV for the production and study of B mesons, and, in addition, it supplies synchrotron radiation for CHESS. The machine has been upgraded for 300 mA operation. It is planned that each beam will be injected in about 5 minutes and that particle beam lifetimes will be several hours. In a cooperative effort, staff members at CHESS and LNS have studied sources in CESR that produce radiation in the user areas. The group has been responsible for the development and realization of new tunnel shielding walls that provide a level of radiation protection from 20 to approx-gt 100 times what was previously available. Our experience has indicated that a major contribution to the environmental radiation is not from photons, but results from neutrons that are generated by particle beam loss in the ring. Neutrons are stopped by inelastic scattering and absorption in thick materials such as heavy concrete. The design for the upgraded walls, the development of a mix for our heavy concrete, and all the concrete casting was done by CHESS and LNS personnel. The concrete incorporates a new material for this application, one that has yielded a significant cost saving in the production of over 200 tons of new wall sections. The material is an artificially enriched iron oxide pellet manufactured in vast quantities from hematite ore for the steel-making industry. Its material and chemical properties (iron and impurity content, strength, size and uniformity) make it an excellent substitute for high grade Brazilian ore, which is commonly used as heavy aggregate in radiation shielding. Its cost is about a third that of the natural ore. The concrete has excellent workability, a 28 day compressive strength exceeding 6000 psi and a density of 220 lbs/cu.ft (3.5 gr/cc). The density is limited by an interesting property of the pellets that is motivated by efficiency in the steel-making application. (Abstract Truncated)

A Reinforcement for Multifunctional Composites for Non-Parasitic Radiation Shielding, Phase II