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Sample records for shielded cell facilities

  1. Shield wall evaluation of hot cell facility for advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Cho, I. J.; Kuk, D. H.; Ko, J. H.; Jung, W. M.; Yoo, G. S.; Lee, E. P.; Park, S. W.

    2002-01-01

    The future hot cell is located in the Irradiated Material Experiment Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). It is β-γ type hot cell that was constructed on the base floor in IMEF building for irradiated material testing. And this hot cell will be used for carrying out the Advanced spent fuel Conditioning Process (ACP). The radiation shielding capability of hot cell should be sufficient to meet the radiation dose requirements in the related regulations. Because the radioactive sources of ACP are expected to be higher than radioactive sources of IMEF design criteria, the future hot cell in current status is unsatisfactory to hot test of ACP. So the shielding analysis of the future hot cell is performed to evaluate shielding ability of concrete shield wall. The shielding analysis included (a) identification of ACP source term; (b) photon source spectrum; (c) shielding analysis by QADS and MCNP-4C; and (d) enhancement of concrete shield wall. In this research, dose rates are obtained according to ACP source, geometry and hot cell shield wall thickness. And the evaluation and reinforcement thickness of the shield wall about future hot cell are concluded

  2. 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

  3. Shielding calculations for the Intense Neutron Source Facility. Final report

    International Nuclear Information System (INIS)

    Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

    1978-06-01

    Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

  4. 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.

  5. 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

  6. Laboratory-scale shielded cell for 252Cf

    International Nuclear Information System (INIS)

    Anderl, R.A.; Cargo, C.H.

    1979-01-01

    A shielded-cell facility for storing and handling remotely up to 2 milligram quantities of unencapsulated 252 Cf has been built in a radiochemistry laboratory at the Test Reactor Area of the Idaho National Engineering Laboratory. Unique features of this facility are its compact bulk radiation shield of borated gypsum and transfer lines which permit the transport of fission product activity from 252 Cf fission sources within the cell to a mass separator and to a fast radiochemistry system in nearby rooms

  7. Shielding experiments for accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2000-06-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  8. Shielding experiments for accelerator facilities

    International Nuclear Information System (INIS)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

    2000-01-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Shielding of Medical Facilities. Shielding Design Considerations for PET-CT Facilities

    International Nuclear Information System (INIS)

    Cruzate, J.A.; Discacciatti, A.P.

    2011-01-01

    The radiological evaluation of a Positron Emission Tomography (PET) 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 evaluation embraces the distributions of rooms, the thickness and physical material of walls, floors and ceilings. This work detail the methodology used for making the assessment of a PET facility design taking into account only radioprotection aspects. The assessment results must be compared to the design requirements established by national regulations in order to determine whether or not, the facility complies with those requirements, both for workers and for members of the public. The analysis presented is useful for both, facility designers and regulators. In addition, some guidelines for improving the shielding design and working procedures are presented in order to help facility designer's job. (authors)

  16. Decommissioning of the Risoe hot cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1992-02-01

    Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, of large contaminated equipment from the concrete cell line and a separate shielded storage facility, and the removal of large contaminated facilities such as out cell parts of a tube transport system between a concrete cell and a lead shielded steel box and a remotely operated Reichert Telatom microscope housed in a lead shielded glove box is described in addition to the initial mapping of radiation levels related to the decontamination of concrete cells. The dose commitment of 17.7 mSv was ascribed to 12 persons in the 2nd half of 1991. The work resulting in these doses was mainly handling of waste together with the frogman entrances in order to repair the in-cell crane and power manipulator. The overall time schedule for the project still appears to be applicable. (AB)

  17. Hot-cell shielding system for high power transmission in DUPIC fuel fabrication

    International Nuclear Information System (INIS)

    Kim, K.; Lee, J.; Park, J.; Yang, M.; Park, H.

    2000-01-01

    This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace. (author)

  18. Shieldings for X-ray radiotherapy facilities calculated by computer

    International Nuclear Information System (INIS)

    Pedrosa, Paulo S.; Farias, Marcos S.; Gavazza, Sergio

    2005-01-01

    This work presents a methodology for calculation of X-ray shielding in facilities of radiotherapy with help of computer. Even today, in Brazil, the calculation of shielding for X-ray radiotherapy is done based on NCRP-49 recommendation establishing a methodology for calculating required to the elaboration of a project of shielding. With regard to high energies, where is necessary the construction of a labyrinth, the NCRP-49 is not very clear, so that in this field, studies were made resulting in an article that proposes a solution to the problem. It was developed a friendly program in Delphi programming language that, through the manual data entry of a basic design of architecture and some parameters, interprets the geometry and calculates the shields of the walls, ceiling and floor of on X-ray radiation therapy facility. As the final product, this program provides a graphical screen on the computer with all the input data and the calculation of shieldings and the calculation memory. The program can be applied in practical implementation of shielding projects for radiotherapy facilities and can be used in a didactic way compared to NCRP-49.

  19. Where have the neutrons gone: A history of the Tower Shielding Facility

    International Nuclear Information System (INIS)

    Muckenthaler, F.J.

    1992-01-01

    In the early 1950's, the concept of the unit shield for the nuclear powered aircraft reactor changed to one of the divided shield concept where the reactor and crew compartment shared the shielding load. Design calculations for the divided shield were being made based on data obtained in studies for the, unit shield. It was believed that these divided shield designs were subject to error, the magnitude of which could not be estimated. This belief led to the design of the Tower Shielding Facility where divided-shield-type measurements could be made without interference from ground or structural scattering. This paper discusses that facility, its reactors, and some chosen experiments from the list of many that were performed at that facility during the past 38 years

  20. The Tower Shielding Facility: Its glorious past

    Energy Technology Data Exchange (ETDEWEB)

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  1. Ford motor company NDE facility shielding design

    International Nuclear Information System (INIS)

    Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

  2. The Tower Shielding Facility: Its glorious past

    International Nuclear Information System (INIS)

    Muckenthaler, F.J.

    1997-01-01

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports

  3. Ford Motor Company NDE facility shielding design.

    Science.gov (United States)

    Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

  4. 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)

  5. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Michael Kruzic

    2007-01-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

  6. New facility shield design criteria

    International Nuclear Information System (INIS)

    Howell, W.P.

    1981-07-01

    The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

  7. Shielding calculations for the design of neutron radiography facility around PARR

    International Nuclear Information System (INIS)

    Ashraf, M.M.; Khan, A.R.

    1989-06-01

    Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)

  8. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

    Higgy, H.R.; Abdel-Rassoul, A.A.

    1983-01-01

    A plan to erect a mechanical shielded hot cell in the process hall of the Radiochemical Laboratory at Inchas is described. The hot cell is designed for safe handling of spent fuel bundles, from the Inchas reactor, and for dismantling and cutting the fuel rods in preparation for subsequent treatment. The biological shielding allows for the safe handling of a total radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an α-tight stainless-steel box, connected to a γ-shielded SAS, through an air-lock containing a movable carriage. The α-box is tightly connected with six dry-storage cavities for adequate storage of the spent fuel bundles. Both the α-box, with the dry-storage cavities, and the SAS are surrounded by 200-mm thick biological lead shielding. The α-box is equipped with two master-slave manipulators, a lead-glass window, a monorail crane and Padirac and Minirag systems. The SAS is equipped with a lead-glass window, tong manipulator, a shielded pit and a mechanism for the entry of the spent fuel bundle. The hot cell is served by adequate ventilation and monitoring systems. (author)

  9. Design and Shielding of Radiotherapy Treatment Facilities; IPEM Report 75, 2nd Edition

    Science.gov (United States)

    Horton, Patrick; Eaton, David

    2017-07-01

    Design and Shielding of Radiotherapy Treatment Facilities provides readers with a single point of reference for protection advice to the construction and modification of radiotherapy facilities. The book assembles a faculty of national and international experts on all modalities including megavoltage and kilovoltage photons, brachytherapy and high-energy particles, and on conventional and Monte Carlo shielding calculations. This book is a comprehensive reference for qualified experts and radiation-shielding designers in radiation physics and also useful to anyone involved in the design of radiotherapy facilities.

  10. Shielded Cells D ampersand D and Dismantlement System Requirements

    International Nuclear Information System (INIS)

    Witherspoon, R.L.

    1995-01-01

    This document describes the basis for the development of the System for Highly Radioactive Equipment Dismantlement or SHRED. It is the result of a thorough investigation into current and past dismantlement practices at shielded cell facilities around the DOE complex. This information has been used to formulate the development requirements for the SHRED

  11. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1993-02-01

    A concise description of the current status (December 31st, 1992) regarding the decommissioning of the hot cell facility at Risoe National Laboratory is given in this periodic report. During the second half of the year 1992, all remaining fissile material and a large amount of contaminated material were removed, major repair work was carried out on the in-cell crane, the shielded storage facility was decontaminated and sealed, iodine filters in the cell ventilation system were removed, remote cleaning was carried out on three concrete cells to radiation levels acceptable for final cleaning by frogmen, and the remaining work schedule was planned. These processes are briefly described. Some breakdowns of older, but vital equipment (i.e. the in-cell crane and the power manipulator) that was taken into extensive use led to a certain amount of delay. The collective radiation doses during this half-year were no higher than under normal operation of the facility, and amounted to 12 man-mSv ascribed to 14 persons. It was concluded that, when removing old epoxy paint in the cells using paint strippers applied by hand, personnel can wear polythene oversuits, although a technique for remote handling has been developed. Tables illustrate measured radiation levels in cells number 1,4,5 and 6, and a diagram describes the shielded storage facility. (AB)

  12. Final flush of the shielded cells melter

    International Nuclear Information System (INIS)

    Marshall, K.M.; Fellinger, T.L.; Harbour, J.R.

    1997-01-01

    A flush of the Savannah River Technology Center (SRTC) Shielded Cells melter was performed after the completion of a campaign to vitrify loaded crystalline silicotitanate (CST) ion exchange medium. The purpose of the flush was to lower levels of radioisotopes accumulated during the campaign and to lower the level of titanium dioxide present in the glass. This in turn would ready the melter for future campaigns involving the Defense Waste Processing Facility (DWPF)

  13. Facility target insert shielding assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-06

    Main objective of this report is to assess the basic shielding requirements for the vertical target insert and retrieval port. We used the baseline design for the vertical target insert in our calculations. The insert sits in the 12”-diameter cylindrical shaft extending from the service alley in the top floor of the facility all the way down to the target location. The target retrieval mechanism is a long rod with the target assembly attached and running the entire length of the vertical shaft. The insert also houses the helium cooling supply and return lines each with 2” diameter. In the present study we focused on calculating the neutron and photon dose rate fields on top of the target insert/retrieval mechanism in the service alley. Additionally, we studied a few prototypical configurations of the shielding layers in the vertical insert as well as on the top.

  14. Shielding Calculations for Positron Emission Tomography - Computed Tomography Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Baasandorj, Khashbayar [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yang, Jeongseon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Integrated PET-CT has been shown to be more accurate for lesion localization and characterization than PET or CT alone, and the results obtained from PET and CT separately and interpreted side by side or following software based fusion of the PET and CT datasets. At the same time, PET-CT scans can result in high patient and staff doses; therefore, careful site planning and shielding of this imaging modality have become challenging issues in the field. In Mongolia, the introduction of PET-CT facilities is currently being considered in many hospitals. Thus, additional regulatory legislation for nuclear and radiation applications is necessary, for example, in regulating licensee processes and ensuring radiation safety during the operations. This paper aims to determine appropriate PET-CT shielding designs using numerical formulas and computer code. Since presently there are no PET-CT facilities in Mongolia, contact was made with radiological staff at the Nuclear Medicine Center of the National Cancer Center of Mongolia (NCCM) to get information about facilities where the introduction of PET-CT is being considered. Well-designed facilities do not require additional shielding, which should help cut down overall costs related to PET-CT installation. According to the results of this study, building barrier thicknesses of the NCCM building is not sufficient to keep radiation dose within the limits.

  15. Accelerator shield design of KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.

    2013-01-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

  16. Introduction of radiation protection and dosimetry in new hot cell facility in research center Rez

    International Nuclear Information System (INIS)

    Svrcula, P.; Petrickova, A.; Srba, O.; Miklos, M.; Svoboda, P.

    2015-01-01

    The purpose of the poster is to present radiation protection and dosimetry in the new hot cell facility being constructed as part of the SUSEN project. The hot cell facility is composed of 10 hot cells and 1 semi-hot cell. All shielding is made from steel, the outer wall shielding has thickness of 500 mm, internal wall between hot cells 300 mm with the possibility to extension to 500 mm. The ceiling shielding has a thickness of 400 mm and the floor shielding is 300 mm wide. Shielded windows allow direct view into the hot cells. Their shielding effect is equivalent to 500 mm of steel. The dimension of the window in the control room is 800 mm x 600 mm with a thickness of 900 mm. All important operating data are collected in the central system of hot cells. The system monitors under-pressure level and temperature in each chamber. If necessary it can directly control the ventilation system. Each hot cell is equipped with dose rate probes. The system also measures and evaluates airborne radioactivity in the building

  17. 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

  18. 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)

  19. Shielding assessment for the proposed HRIBF upgrade to the National ISOL Facility

    International Nuclear Information System (INIS)

    Slater, C.O.; Olsen, D.K.; Johnson, J.O.; Lillie, R.A.; Gabriel, T.A.

    1997-04-01

    An upgrade of the existing ORNL Holifield Radioactive Ion Beam Facility (HRIBF) to the National Radioactive Ion Beam Isotope Separator On Line (RIB ISOL) Facility is being proposed. Part of the upgrade involves increasing the source proton energy and current, resulting in more intense, higher energy radiation. Shielding requirements for the proposed upgrade to the HRIBF have been assessed with respect to weight, space, and dose-rate constraints. Shielding assessments were made for operating, shutdown, and accident conditions. The results indicate reasonable shielding solutions for the target room except for the marginal dose rate on the roof. Shielding requirements in the target room were greatly reduced by decisions to move the target to a more interior room and to direct the proton beam downward into the target. A slightly more difficult shielding problem arises for proton beam extraction losses from the cyclotron. Here, the assumed isotropic beam losses (hence, neutron emissions) mean higher roof dose rates than those over the target room unless substantial localized shielding is placed over the cyclotron. Shutdown dose rates were found to present no problems. While dose rates through the sides of the facility during accident conditions will probably satisfy the accident dose-rate constraints, dose rates above the roof will be well above the constraints unless a solution is devised to shield the locations where beam losses are likely to occur. Ground activation analysis was postponed for this study

  20. Shielding analysis of high level waste water storage facilities using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Yabuta, Naohiro [Mitsubishi Research Inst., Inc., Tokyo (Japan)

    2001-01-01

    The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)

  1. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1994-06-01

    Nuclear fuels have been handled and examined after irradiation by physical and chemical techniques, and radiotherapy sources, mainly 60 Co, have been produced at Risoe National Laboratory since 1964. The aims of decommissioning (during 1990-94, at IAEA Stage 2 level for reactors) were to obtain safe conditions for the remaining parts of the facility and to produce clean space areas for new projects. The facility comprises 6 concrete cells, several lead-shielded steel cells, glove boxes, shielded storage for waste, a remotely operated optical microscope, a frogman area for personnel access to the concrete cells, a decontamination facility, workshops and safety installations. All steel cells, glove boxes and the microscope were emptied and removed. The concrete cells were emptied of fissile material, scientific equipment, general tools and scrap. Decontamination should facilitate waste packing and reduce amount of waste to be stored temporarily at the Risoe waste treatment facility together with highly active waste. The concrete cells were cleaned remotely by wiping, hot spot removal, by mechanical means and vacuum cleaning. The interiors of 2 cells were decontaminated by high pressure water jetting. All master-slave manipulators and part of the contaminated ventilation system at the cells were removed. The cells are left in a non-ventilated state, connected to the atmosphere by an absolute filter. The main contaminants measured before cell closure were 60 Co, 137 Cs and alpha-emitters. Dismantling, decontamination waste disposal and received doses are described. Simple techniques involving low doses were found to be very effective. (AB)

  2. The Spallation Neutron Source (SNS) conceptual design shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Odano, N.; Lillie, R.A.

    1998-03-01

    The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

  3. Experience with a servo-hydraulic mechanical testing machine installed in a new shielded active facility at Windscale Nuclear Power Development Laboratories

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.; Gravenor, J.G.; Rhodes, D.

    1982-03-01

    An Instron model 1273 servo-hydraulic machine has been installed within a lead-shielded cell at Windscale in order to provide a facility capable of performing a wide range of mechanical tests on nuclear reactor structural materials and fuel assembly components. This particular type of machine was chosen because it has design features associated with the load frame, location of the actuator and adjustment and clamping of the cross-head that are especially well suited to remote operation within a shielded cell. The design of the testing facility is described and the programmes of work that have been completed over the past 11/2 years of operation are reviewed. (author)

  4. Gamma-ray shielding design and performance test of WASTEF

    International Nuclear Information System (INIS)

    Matsumoto, Seiichiro; Aoyama, Saburo; Tashiro, Shingo; Nagai, Shiro

    1984-06-01

    The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 10 4 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 10 6 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)

  5. Characterisation study of radionuclides in Hot Cell Facility

    International Nuclear Information System (INIS)

    Ghare, P.T.; Rath, D.P.; Govalkar, Atul; Mukherjee, Govinda; AniIKumar, S.; Yadav, R.K.B.; Mallik, G.K.

    2016-01-01

    Examination of different types of experimental as well as power reactor irradiated fuels and validation of fuel modeling codes is carried out in general Hot cell facility. The Hot cell facility has six concrete shielded hot cells, capable of handling radioactivity varying from 3.7 TBQ to 3700 TBq gamma activity. The facility was augmented with two hot cells having designed capacity to handle radioactivity of 9250 TBQ of equivalent activity of 60 Co. The study of characterization of various radionuclides present inside the hot cell of PIE facility was taken up. This study will help in providing valuable inputs for various radiological safety parameters to keep personnel exposure to ALARA level as per the mandate of radiation safety program

  6. Optimization of the National Ignition Facility primary shield design

    International Nuclear Information System (INIS)

    Annese, C.E.; Watkins, E.F.; Greenspan, E.; Miller, W.F.

    1993-10-01

    Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries

  7. Astronaut Ronald Sega with Wake Shield Facility on test stand at JSC

    Science.gov (United States)

    1991-01-01

    The Wake Shield Facility is displayed on a test stand at JSC. Astronaut Ronald M. Sega, mission specialist for STS-60, is seen with the facility during a break in testing in the acoustic and vibration facility at JSC.

  8. Refurbishment of an Analytical Laboratory Hot Cell Facility

    International Nuclear Information System (INIS)

    Rosenberg, K.; Henslee, S.P.; Michelbacher, J.A.; Coleman, R.M.

    1997-01-01

    An Analytical Laboratory Hot Cell (ALHC) Facility at Argonne National Laboratory-West (ANL-W) was in service for nearly thirty years. In order to comply with DOE regulations governing such facilities and meet ANL-W programmatic requirements, a major refurbishment effort was undertaken. All penetrations within the facility were sealed; the ventilation system was redesigned, upgraded and replaced; the manipulators were replaced; the hot cell windows were removed, refurbished, and reinstalled; all hot cell utilities were replaced; a lead-shielded glovebox housing an Inductively Coupled Plasma - Atomic Emission Spectrometer (ICP-AES) System was interfaced with the hot cells, and a new CO2 fire suppression system and other ALHC support equipment were installed

  9. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Souza, Gregorio Soares de

    2011-01-01

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  10. Electron accelerator shielding design of KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

    2016-06-15

    The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

  11. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Michael R, Kruzic

    2008-01-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  12. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Energy Technology Data Exchange (ETDEWEB)

    Michael R. Kruzic

    2008-06-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  13. Lessons from shielding retrofits at the LAMPF/LANSCE/PSR accelerator, beam lines and target facilities

    International Nuclear Information System (INIS)

    Macek, R.J.

    1994-01-01

    The experience in the past 7 years to improve the shielding and radiation control systems at the Los Alamos Meson Physics Facility (LAMPF) and the Manuel Lujan Jr. Neutron Scattering Center (LANSCE) provides important lessons for the design of radiation control systems at future, high beam power proton accelerator facilities. Major issues confronted and insight gained in developing shielding criteria and in the use of radiation interlocks are discussed. For accelerators and beam lines requiring hands-on-maintenance, our experience suggests that shielding criteria based on accident scenarios will be more demanding than criteria based on routinely encountered beam losses. Specification and analysis of the appropriate design basis accident become all important. Mitigation by active protection systems of the consequences of potential, but severe, prompt radiation accidents has been advocated as an alternate choice to shielding retrofits for risk management at both facilities. Acceptance of active protection systems has proven elusive primarily because of the difficulty in providing convincing proof that failure of active systems (to mitigate the accident) is incredible. Results from extensive shielding assessment studies are presented including data from experimental beam spill tests, comparisons with model estimates, and evidence bearing on the limitations of line-of-sight attenuation models in complex geometries. The scope and significant characteristics of major shielding retrofit projects at the LAMPF site are illustrated by the project to improve the shielding beneath a road over a multiuse, high-intensity beam line (Line D)

  14. Conservative method for determination of material thickness used in shielding of veterinary facilities

    International Nuclear Information System (INIS)

    Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

    2014-01-01

    For determination of an effective method for shielding of veterinary rooms, was provided shielding methods generally used in rooms which works with X-ray production and radiotherapy. Every calculation procedure is based in traditional variables used to transmission calculation. The thickness of the materials used for primary and secondary shieldings are obtained to respect the limits set by the Brazilian National Nuclear Energy Commission (CNEN). This work presents the development of a computer code in order to serve as a practical tool for determining rapid and effective materials and their thicknesses to shield veterinary facilities. The code determines transmission values of the shieldings and compares them with data from transmission 'maps' provided by NCRP-148 report. These 'maps' were added to the algorithm through interpolation techniques of curves of materials used for shielding. Each interpolation generates about 1,000,000 points that are used to generate a new curve. The new curve is subjected to regression techniques, which makes possible to obtain nine degree polynomial, and exponential equations. These equations whose variables consist of transmission of values, enable trace all the points of this curve with high precision. The data obtained from the algorithm were satisfactory with official data presented by the National Council of Radiation Protection and Measurements (NCRP) and can contribute as a practical tool for verification of shielding of veterinary facilities that require using Radiotherapy techniques and X-ray production

  15. Neutron streaming analysis for shield design of FMIT Facility

    International Nuclear Information System (INIS)

    Carter, L.L.

    1980-12-01

    Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

  16. Operating manual for the Tower Shielding Facility

    International Nuclear Information System (INIS)

    1985-12-01

    This manual provides information necessary to operate and perform maintenance on the reactor systems and all equipment or systems which can affect their operation or the safety of personnel at the Tower Shielding Facility. The first four chapters consist of introductory and descriptive material of benefit to personnel in training, the qualifications required for training, the responsibilities of the personnel in the organization, and the procedures for reviewing proposed experiments. Chapter 8, Emergency Procedures, is also a necessary part of the indoctrination of personnel. The procedures for operation of the Tower Shielding Reactor (TSR-II), its water cooling system, and the main tower hoists are outlined in Chapters 5, 6, and 7. The Technical Specification surveillance requirements for the TSR-II are summarized in Chapter 9. The maintenance and calibration schedule is spelled out in Chapter 10. The procedures for assembly and disassembly of the TSR-II are outlined in Chapter 11

  17. Progress on establishing guidelines for National Ignition Facility (NIF) experiments to extend debris shield lifetime

    International Nuclear Information System (INIS)

    Tobin, M.; Eder, D.; Braun, D.; MacGowan, B.

    2002-01-01

    The survivability of the debris shields on the National Ignition Facility (NIF) are a key factor for the affordable operation of the facility. The improvements required over Nova debris shields are described. Estimates of debris shield lifetimes in the presence of target emissions with 4-8 J/cm 2 laser fluences indicate lifetimes that may contribute unacceptably to operations costs for NIF. We are developing detailed suggested guidance for target and experiment designers for NIF to assist in minimizing the damage to, and therefore the cost of, maintaining NIF debris shields. The guidance suggests a target mass quantity that as particulate on the debris shields (300 mg) may be within current operating budgets. It also suggests the amount of material that should become shrapnel on a shot (10 mg). Finally, it suggests the level of non-volatile residue (NVR) that would threaten the sol-gel coatings on the debris shields (1 μg/cm 2 ). We review the experimentation on the Nova chamber that included measuring quantities of particulate on debris shields by element and capturing shrapnel pieces in aerogel samples mounted in the chamber. We also describe computations of X-ray emissions from a likely NIF target and the associated ablation expected from this X-ray exposure on supporting target hardware. We describe progress in assessing the benefits of a pre-shield and the possible impact on the guidance for target experiments on NIF. Plans for possible experimentation on Omega and other facilities to improve our understanding of target emissions and their impacts are discussed. Our discussion of planned future work provides a forum to invite possible collaboration with the IFE community

  18. Benchmarking study and its application for shielding analysis of large accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk [POSTECH, Pohang (Korea, Republic of)

    2015-10-15

    Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes.

  19. Benchmarking study and its application for shielding analysis of large accelerator facilities

    International Nuclear Information System (INIS)

    Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk

    2015-01-01

    Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes

  20. 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

  1. 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

  2. Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF-11 Workshop Proceedings Report

    International Nuclear Information System (INIS)

    2013-01-01

    Particle accelerators have evolved over the last decades from simple devices to powerful machines. In recent years, new technological and research applications have helped to define requirements while the number of accelerator facilities in operation, being commissioned, designed or planned has grown significantly. Their parameters, which include the beam energy, currents and intensities, and target composition, can vary widely, giving rise to new radiation shielding issues and challenges. Particle accelerators must be operated in safe ways to protect operators, the public and the environment. As the design and use of these facilities evolve, so must the analytical methods used in the safety analyses. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and irradiation targets. They also evaluate progress in the development of modelling methods used to assess the effectiveness of such shielding as part of safety analyses. The transport of radiation through shielding materials is a major consideration in the safety design studies of nuclear power plants, and the modelling techniques used may be applied to many other types of scientific and technological facilities. Accelerator and irradiation facilities represent a key capability in R and D, medical and industrial infrastructures, and they can be used in a wide range of scientific, medical and industrial applications. High-energy ion accelerators, for example, are now used not only in fundamental research, such as the search for new super-heavy nuclei, but also for therapy as part of cancer treatment. While the energy of the incident particles on the shielding of these facilities may be much higher than those found in nuclear power plants, much of the physics associated with the behaviour of the secondary particles produced is similar, as are the computer modelling techniques used to quantify key safety design parameters, such as radiation dose and activation levels

  3. Shield evaluation and validation for design and operation of facility for treatment of legacy Intermediate Level Radioactive Liquid Waste (ILW)

    International Nuclear Information System (INIS)

    Deepa, A.; Jakhete, A.P.; Rathish, K.R.; Saroj, S.K.; Patel, H.S.; Gopalakrishnan, R.K.; Gangadharan, Anand; Singh, Neelima

    2014-01-01

    An ion exchange treatment facility has been commissioned at PRIX facility, for the treatment of legacy ILW generated at reprocessing plant, Trombay. The treatment system is based on the deployment of selective sorbents for removal of cesium and strontium from ILW. Activity concentration due to beta emitters likely to be processed is of the order of 111-1850 MBq/l. Dose rates in different areas of the facility were evaluated using shielding code and design input. Present work give details of the comparison of dose rates estimated and dose rates measured at various stages of the processing of ILW. At PRIX, the ILW treatment system comprises of shielded IX columns (two cesium and one strontium) housed in a MS cubicle the process lines inlet and outlet of IX treatment system and effluent storage tanks. The MS cubicle, prefilter and piping are housed in a process cell of 500 mm concrete shielding. Effluent storage tanks are outside processing building. Theoretical assessment of expected dose rates were carried out prior to installation of various systems in different areas of PRIX. Dose rate on IX column and MS cubicle for a maximum inventory of 3.7x10 7 MBq of 137 Cs and its contribution in operating gallery was estimated

  4. Upgrades of Hanford Engineering Development Laboratory hot cell facilities

    International Nuclear Information System (INIS)

    Daubert, R.L.; DesChane, D.J.

    1987-01-01

    The Hanford Engineering Development Laboratory operates the 327 Postirradiation Testing Laboratory (PITL) and the 324 Shielded Materials Facility (SMF). These hot cell facilities provide diverse capabilities for the postirradiation examination and testing of irradiated reactor fuels and materials. The primary function of these facilities is to determine failure mechanisms and effects of irradiation on physical and mechanical properties of reactor components. The purpose of this paper is to review major equipment and facility upgrades that enhance customer satisfaction and broaden the engineering capabilities for more diversified programs. These facility and system upgrades are providing higher quality remote nondestructive and destructive examination services with increased productivity, operator comfort, and customer satisfaction

  5. Neutron/photon/electron shielding study for a laser-fusion facility

    International Nuclear Information System (INIS)

    Thompson, W.L.

    1977-01-01

    A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

  6. Measurements and FLUKA Simulations of Bismuth, Aluminium and Indium Activation at the upgraded CERN Shielding Benchmark Facility (CSBF)

    Science.gov (United States)

    Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.; Yashima, H.

    2018-06-01

    The CERN High energy AcceleRator Mixed field (CHARM) facility is situated in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5·1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7·1010 protons per second. The extracted proton beam impacts on a cylindrical copper target. The shielding of the CHARM facility includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target that allows deep shielding penetration benchmark studies of various shielding materials. This facility has been significantly upgraded during the extended technical stop at the beginning of 2016. It consists now of 40 cm of cast iron shielding, a 200 cm long removable sample holder concrete block with 3 inserts for activation samples, a material test location that is used for the measurement of the attenuation length for different shielding materials as well as for sample activation at different thicknesses of the shielding materials. Activation samples of bismuth, aluminium and indium were placed in the CSBF in September 2016 to characterize the upgraded version of the CSBF. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields of bismuth isotopes (206 Bi, 205 Bi, 204 Bi, 203 Bi, 202 Bi, 201 Bi) from 209 Bi, 24 Na from 27 Al and 115 m I from 115 I for these samples. The production yields estimated by FLUKA Monte Carlo simulations are compared to the production yields obtained from γ-spectroscopy measurements of the samples taking the beam intensity profile into account. The agreement between FLUKA predictions and γ-spectroscopy measurements for the production yields is at a level of a factor of 2.

  7. Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

    International Nuclear Information System (INIS)

    Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

    2004-01-01

    To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

  8. ORNL shielded facilities capable of remote handling of highly radioactive beta--gamma emitting materials

    International Nuclear Information System (INIS)

    Whitson, W.R.

    1977-09-01

    A survey of ORNL facilities having adequate shielding and containment for the remote handling of experimental quantities of highly radioactive beta-gamma emitting materials is summarized. Portions of the detailed descriptions of these facilities previously published in ORNL/TM-1268 are still valid and are repeated

  9. SATIF-2 shielding aspects of accelerators, targets and irradiation facilities

    International Nuclear Information System (INIS)

    1995-01-01

    Particle accelerators have evolved over the last 50 years from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today they cover a wide range of applications, from television and computer displays in households to the investigation of the origin and structure of matter. It has become common practice to use them for material science and medical applications. In recent years, requirements from new technological and research applications have emerged, such as increased particle beams intensities, higher flexibility, etc., giving rise to new radiation shielding aspects and problems. These proceedings review recent progress in radiation shielding of accelerator facilities, and evaluate advancements with respect to international co-operation in this field

  10. Design of auxiliary shield for remote controlled metallographic microscope

    International Nuclear Information System (INIS)

    Matsui, Hiroki; Okamoto, Hisato

    2014-06-01

    The remote controlled optical microscope installed in the lead cell at the Reactor Fuel Examination Facility (RFEF) in Japan Atomic Energy Agency (JAEA) has been upgraded to a higher performance unit to study the effect of the microstructural evolution in clad material on the high burn-up fuel behavior under the accident condition. The optical pass of the new microscope requires a new through hole in the shielding lead wall of the cell. To meet safety regulations, auxiliary lead shieldings were designed to cover the lost shielding function of the cell wall. Particle and Heavy Ion Transport Code System (PHITS) was used to calculate and determine the shape and setting positions of the shielding unit. Seismic assessments of the unit were also performed. (author)

  11. Design of a PET/CT facility considering the shielding calculation in accordance with AAPM TG-108

    International Nuclear Information System (INIS)

    Guevara R, V. Y.; Romero C, N.; Berrocal T, M.

    2014-08-01

    A Positron Emission Tomography / Computed Tomography facility may require protection barriers on floor, ceiling and walls, because the patient becomes a radioactive source that emits photons of 0.511 MeV, after having received a radiopharmaceutical, usually F-18 fluorodeoxyglucose (F-18 FDG). This work has as objective to propose the design of a PET/CT facility, taking into account technical and radiation protection considerations applied internationally, and also develop the necessary shielding for such installation by applying as published by the American Association of Physicists in Medicine Task Group Report 108. A shielding spreadsheet in Excel program was developed with reference to the recommendations of the AAPM TG - 08, to determine the shielding required for the walls, floor and ceiling. For fixing the radiation levels in the shielding calculation has been considered the actual restrictions for the occupationally exposed personnel (100 μSv/week) as well as the people in general (20 μSv/ week). The radiopharmaceutical used as a reference for the shielding calculation was the F-18 FDG. With the assistance of an architectural plan were determined distances from potential sources of radiation in facility (uptake and image acquisition living rooms) to points of interest around them. Finally the thickness of the protective barriers in lead and concrete necessary to achieve the established radiation levels were calculated and these results were stored in a table. This paper shows that technical aspects considered in the design of the installation and environments distribution can improve work processes within the PET/CT facility, consequently resulting in a reduction of the dose levels for people in general. (author)

  12. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    Gohar, Y.

    1985-01-01

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m 2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

  13. Important changes in medical x-ray imaging facility shielding design methodology. A brief summary of recommendations in NCRP Report No. 147

    International Nuclear Information System (INIS)

    Archer, Benjamin R.; Gray, Joel E.

    2005-01-01

    The recently published Report No. 147 of The National Council on Radiation Protection and Measurements entitled 'Structural shielding design for medical x-ray imaging facilities' provides an update of shielding recommendations for x rays used for medical imaging. The goal of this report is to ensure that the shielding in these facilities limits radiation exposures to employees and members of the public to acceptable levels. Board certified medical and health physicists, as defined in this report, are the 'qualified experts' who are competent to design radiation shielding for these facilities. As such, physicists must be aware of the new technical information and the changes from previous reports that Report No. 147 supersedes. In this article we summarize the new data, models and recommendations for the design of radiation barriers in medical imaging facilities that are presented in Report No. 147

  14. Design of the magnetized muon shield for the prompt-neutrino facility

    International Nuclear Information System (INIS)

    Baltay, C.; Bosek, N.; Couch, J.

    1982-01-01

    The main technical challenge in the design of the prompt neutrino beam is the magnetized muon shield. Two satisfactory alternate designs have been developed for such a shield during this past year and the background muon fluxes have been calculated by three independent programs at Columbia, Fermilab, and MIT. The background muon fluxes have been calculated to be satisfactory in all of the detectors that might use the beam. In Section III of this report we describe in detail the three Monte Carlo programs used in these calculations. In Section IV we give the details of the flux calculations for the E-613 shield and the comparisons with the observed fluxes with various configurations of that shield. In Section V we describe the designs that have been developed for the neutrino area shield. In Section VI we discuss the problem of proton beam transport losses and the associated muon fluxes. Finally, in Section VII a comparison of the two solutions is made which covers cost, effectiveness, schedule and responsiveness to future unknowns. We conclude that there are not overwhelming reasons for the choice of one design over the other. However, for a variety of secondary reasons the superconducting design offers advantages. We therefore propose the construction of the prompt neutrino facility with the superconducting magnet design

  15. 3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, Masaya; Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

  16. Evaluation of bulk shield for the JHP facilities

    International Nuclear Information System (INIS)

    Uwamino, Yoshitomo; Shibata, Tokushi

    1991-01-01

    In the Japanese Hadron Project (JHP), a 1-GeV 200-μA proton beam will be handled, and the radiation shield of the facility will be very massive concrete and iron lump. Since the constructing cost is strongly affected by the shielding design, the design must be severely performed. The neutron yield in thin targets and a copper beam dump was calculated by the HETC-KFA-2 Monte Carlo code. For the evaluation of the calculational accuracy, the calculational results were compared with the experimental data by Cierjacks and Raupp. The calculated result of heavy element agreed well with the experiment at a low energy region, E n n >100 MeV) of 90 deg close to the calculated one of about 60 deg in the absolute value. The high energy neutron transport in a 5-m-thick iron slab and in an 8-m-thick ordinary concrete slab was calculated with the HETC code and also with the discrete ordinates transport code, ANISN. In the ANISN calculation, the DLC-87/HILO and the DLC-128/LAHIMAC group cross sections were used. The ANISN calculation with the LAHIMAC cross sections gave strong underestimation compared with the HETC calculation. The difference of the shielding lengths calculated by the HETC code and by the ANISN code used with the HILO cross sections was smaller than 6% for the both iron and concrete cases. (author)

  17. Preliminary Shielding Assessment for the IFF System in the RAON Heavy-ion Facility

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Lee, Youngouk; Kim, Jong Won; Kim, Mijung

    2014-01-01

    A heavy-ion accelerator facility is under a development in Korea to use in the basic science research and various application areas. In this facility, the In-Flight Fragment (IFF) target and isotope separator has been designed to produce various isotopes and transport the interesting isotopes into the experimental rooms. In this work, preliminary radiation shielding assessment was performed for the IFF target room

  18. Bulk-shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

    1982-07-01

    The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

  19. 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

  20. A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

    International Nuclear Information System (INIS)

    Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

    2013-01-01

    In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

  1. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J S; Choi, J W; Go, W I; Kim, H D; Song, K C; Jeong, I H; Park, H S; Im, C S; Lee, H M; Moon, K H; Hong, K P; Lee, K S; Suh, K S; Kim, E K; Min, D K; Lee, J C; Chun, Y B; Paik, S Y; Lee, E P; Yoo, G S; Kim, Y S; Park, J C

    1997-09-01

    In the early stage of the project, a comprehensive survey was conducted to identify the feasibility of using available facilities and of interface between those facilities. It was found out that the shielded cell M6 interface between those facilities. It was found out that the shielded cell M6 of IMEF could be used for the main process experiments of DUPIC fuel fabrication in regard to space adequacy, material flow, equipment layout, etc. Based on such examination, a suitable adapter system for material transfer around the M6 cell was engineered. Regarding the PIEF facility, where spent PWR fuel assemblies are stored in an annex pool, disassembly devices in the pool are retrofitted and spent fuel rod cutting and shipping system to the IMEF are designed and built. For acquisition of casks for radioactive material transport between the facilities, some adaptive refurbishment was applied to the available cask (Padirac) based on extensive analysis on safety requirements. A mockup test facility was newly acquired for remote test of DUPIC fuel fabrication process equipment prior to installation in the M6 cell of the IMEF facility. (author). 157 refs., 57 tabs., 65 figs.

  2. DUPIC facility engineering

    International Nuclear Information System (INIS)

    Lee, J. S.; Choi, J. W.; Go, W. I.; Kim, H. D.; Song, K. C.; Jeong, I. H.; Park, H. S.; Im, C. S.; Lee, H. M.; Moon, K. H.; Hong, K. P.; Lee, K. S.; Suh, K. S.; Kim, E. K.; Min, D. K.; Lee, J. C.; Chun, Y. B.; Paik, S. Y.; Lee, E. P.; Yoo, G. S.; Kim, Y. S.; Park, J. C.

    1997-09-01

    In the early stage of the project, a comprehensive survey was conducted to identify the feasibility of using available facilities and of interface between those facilities. It was found out that the shielded cell M6 interface between those facilities. It was found out that the shielded cell M6 of IMEF could be used for the main process experiments of DUPIC fuel fabrication in regard to space adequacy, material flow, equipment layout, etc. Based on such examination, a suitable adapter system for material transfer around the M6 cell was engineered. Regarding the PIEF facility, where spent PWR fuel assemblies are stored in an annex pool, disassembly devices in the pool are retrofitted and spent fuel rod cutting and shipping system to the IMEF are designed and built. For acquisition of casks for radioactive material transport between the facilities, some adaptive refurbishment was applied to the available cask (Padirac) based on extensive analysis on safety requirements. A mockup test facility was newly acquired for remote test of DUPIC fuel fabrication process equipment prior to installation in the M6 cell of the IMEF facility. (author). 157 refs., 57 tabs., 65 figs

  3. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range

    International Nuclear Information System (INIS)

    Hanslik, R.

    2006-08-01

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

  4. Preliminary shielding estimates for the proposed Oak Ridge National Laboratory (ORNL) Radioactive Ion Beam Facility (RIBF)

    International Nuclear Information System (INIS)

    Johnson, J.O.; Gabriel, T.A.; Lillie, R.A.

    1996-01-01

    The Oak Ridge National Laboratory (ORNL) has proposed designing and implementing a new target-ion source for production and injection of negative radioactive ion beams into the Hollifield tandem accelerator. This new facility, referred to as the Radioactive Ion Beam Facility (RIBF), will primarily be used to advance the scientific communities' capabilities for performing state-of-the-art cross-section measurements. Beams of protons or other light, stable ions from the Oak Ridge Isochronous Cyclotron (ORIC) will be stopped in the RIBF target ion source and the resulting radioactive atoms will be ionized, charge exchanged, accelerated, and injected into the tandem accelerator. The ORIC currently operates with proton energies up to 60 MeV and beam currents up to 100 microamps with a maximum beam power less than 2.0 kW. The proposed RIBF will require upgrading the ORIC to generate proton energies up to 200 MeV and beam currents up to 200 microamps for optimum performance. This report summarizes the results of a preliminary one-dimensional shielding analysis of the proposed upgrade to the ORIC and design of the RIBF. The principal objective of the shielding analysis was to determine the feasibility of such an upgrade with respect to existing shielding from the facility structure, and additional shielding requirements for the 200 MeV ORIC machine and RIBF target room

  5. Radiation Attenuation and Stability of ClearView Radiation Shielding TM-A Transparent Liquid High Radiation Shield.

    Science.gov (United States)

    Bakshi, Jayeesh

    2018-04-01

    Radiation exposure is a limiting factor to work in sensitive environments seen in nuclear power and test reactors, medical isotope production facilities, spent fuel handling, etc. The established choice for high radiation shielding is lead (Pb), which is toxic, heavy, and abidance by RoHS. Concrete, leaded (Pb) bricks are used as construction materials in nuclear facilities, vaults, and hot cells for radioisotope production. Existing transparent shielding such as leaded glass provides minimal shielding attenuation in radiotherapy procedures, which in some cases is not sufficient. To make working in radioactive environments more practicable while resolving the lead (Pb) issue, a transparent, lightweight, liquid, and lead-free high radiation shield-ClearView Radiation Shielding-(Radium Incorporated, 463 Dinwiddie Ave, Waynesboro, VA). was developed. This paper presents the motivation for developing ClearView, characterization of certain aspects of its use and performance, and its specific attenuation testing. Gamma attenuation testing was done using a 1.11 × 10 Bq Co source and ANSI/HPS-N 13.11 standard. Transparency with increasing thickness, time stability of liquid state, measurements of physical properties, and performance in freezing temperatures are reported. This paper also presents a comparison of ClearView with existing radiation shields. Excerpts from LaSalle nuclear power plant are included, giving additional validation. Results demonstrated and strengthened the expected performance of ClearView as a radiation shield. Due to the proprietary nature of the work, some information is withheld.

  6. Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

    International Nuclear Information System (INIS)

    OSCAR, DEBBY S.; WALKER, SHARON ANN; HUNTER, REGINA LEE; WALKER, CHERYL A.

    1999-01-01

    The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2

  7. Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

    Energy Technology Data Exchange (ETDEWEB)

    OSCAR,DEBBY S.; WALKER,SHARON ANN; HUNTER,REGINA LEE; WALKER,CHERYL A.

    1999-12-01

    The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2.

  8. Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

    CERN Document Server

    Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

    2002-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

  9. 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)

  10. SU-E-T-400: Evaluation of Shielding and Activation at Two Pencil Beam Scanning Proton Facilities

    International Nuclear Information System (INIS)

    Remmes, N; Mundy, D; Classic, K; Beltran, C; Kruse, J; Herman, M; Stoker, J; Nelson, K; Bues, M

    2015-01-01

    Purpose: To verify acceptably low dose levels around two newly constructed identical pencil beam scanning proton therapy facilities and to evaluate accuracy of pre-construction shielding calculations. Methods: Dose measurements were taken at select points of interest using a WENDI-2 style wide-energy neutron detector. Measurements were compared to pre-construction shielding calculations. Radiation badges with neutron dose measurement capabilities were worn by personnel and also placed at points throughout the facilities. Seven neutron and gamma detectors were permanently installed throughout the facility, continuously logging data. Potential activation hazards have also been investigated. Dose rates near water tanks immediately after prolonged irradiation have been measured. Equipment inside the treatment room and accelerator vault has been surveyed and/or wipe tested. Air filters from air handling units, sticky mats placed outside of the accelerator vault, and water samples from the magnet cooling water loops have also been tested. Results: All radiation badges have been returned with readings below the reporting minimum. Measurements of mats, air filters, cooling water, wipe tests and surveys of equipment that has not been placed in the beam have all come back at background levels. All survey measurements show the analytical shielding calculations to be conservative by at least a factor of 2. No anomalous events have been identified by the building radiation monitoring system. Measurements of dose rates close to scanning water tanks have shown dose rates of approximately 10 mrem/hr with a half-life less than 5 minutes. Measurements around the accelerator show some areas with dose rates slightly higher than 10 mrem/hr. Conclusion: The shielding design is shown to be adequate. Measured dose rates are below those predicted by shielding calculations. Activation hazards are minimal except in certain very well defined areas within the accelerator vault and for objects

  11. Shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

    1983-03-01

    The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

  12. Functional components for a design strategy: Hot cell shielding in the high reliability safeguards methodology

    Energy Technology Data Exchange (ETDEWEB)

    Borrelli, R.A., E-mail: rborrelli@uidaho.edu

    2016-08-15

    The high reliability safeguards (HRS) methodology has been established for the safeguardability of advanced nuclear energy systems (NESs). HRS is being developed in order to integrate safety, security, and safeguards concerns, while also optimizing these with operational goals for facilities that handle special nuclear material (SNM). Currently, a commercial pyroprocessing facility is used as an example system. One of the goals in the HRS methodology is to apply intrinsic features of the system to a design strategy. This current study investigates the thickness of the hot cell walls that could adequately shield processed materials. This is an important design consideration that carries implications regarding the formation of material balance areas, the location of key measurement points, and material flow in the facility.

  13. Analysis of a shield design for a DT neutron generator test facility.

    Science.gov (United States)

    Chichester, D L; Pierce, G D

    2007-10-01

    Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

  14. Double-layer neutron shield design as neutron shielding application

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

  15. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Pace, J.V. III.

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  16. Measurements and FLUKA simulations of bismuth and aluminium activation at the CERN Shielding Benchmark Facility (CSBF)

    Science.gov (United States)

    Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.

    2018-03-01

    The CERN High Energy AcceleRator Mixed field facility (CHARM) is located in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5 ṡ1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7 ṡ1010 p/s that then impacts on the CHARM target. The shielding of the CHARM facility also includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target. This facility consists of 80 cm of cast iron and 360 cm of concrete with barite concrete in some places. Activation samples of bismuth and aluminium were placed in the CSBF and in the CHARM access corridor in July 2015. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields for these samples. The results estimated by FLUKA Monte Carlo simulations are compared to activation measurements of these samples. The comparison between FLUKA simulations and the measured values from γ-spectrometry gives an agreement better than a factor of 2.

  17. Measurement of 36Cl induced in shielding concrete of various accelerator facilities

    International Nuclear Information System (INIS)

    Bessho, K.; Matsumura, H.; Matsuhiro, T.

    2003-01-01

    The concentrations of 36 Cl induced in shielding concrete of the various accelerators has been measured by accelerator mass spectrometry. For three kinds of accelerator facilities, SF cyclotron (Center for Nuclear Study, the University of Tokyo), 300 MeV electron LINAC (Laboratory of Nuclear Science, Tohoku University), and 12 GeV proton synchrotron (High Energy Accelerator Research Organization), the depth profiles of 36 Cl/ 35 Cl ratios in concrete samples near the beam lines were analyzed. The depth profiles of 36 Cl/ 35 Cl are consistent with those of the radioactive concentrations of 152 Eu and 60 Co, which are formed by thermal neutron capture reactions. These results imply that 36 Cl formed in shielding concrete of these accelerators is mainly produced by thermal neutron capture of 35 Cl. The maximum 36 Cl/ 35 Cl ratio of 3x10 -8 (300 MeV electron LINAC, depth of 8 cm) corresponds to the specific radioactivity of 2x10 -3 Bq/g, which is not serious for radioactive waste management in reconstruction or decommissioning of accelerator facilities, compared with specific radioactivity of 3 H, 152 Eu and 60 Co. (author)

  18. SHIELD 1.0: development of a shielding calculator program in diagnostic radiology

    International Nuclear Information System (INIS)

    Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da; Friedrich, Barbara Q.; Silva, Ana Maria Marques da

    2013-01-01

    In shielding calculation of radiological facilities, several parameters are required, such as occupancy, use factor, number of patients, source-barrier distance, area type (controlled and uncontrolled), radiation (primary or secondary) and material used in the barrier. The shielding design optimization requires a review of several options about the physical facility design and, mainly, the achievement of the best cost-benefit relationship for the shielding material. To facilitate the development of this kind of design, a program to calculate the shielding in diagnostic radiology was implemented, based on data and limits established by National Council on Radiation Protection and Measurements (NCRP) 147 and SVS-MS 453/98. The program was developed in C⌗ language, and presents a graphical interface for user data input and reporting capabilities. The module initially implemented, called SHIELD 1.0, refers to calculating barriers for conventional X-ray rooms. The program validation was performed by the comparison with the results of examples of shielding calculations presented in NCRP 147.

  19. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-10-15

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  20. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    International Nuclear Information System (INIS)

    Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-01-01

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  1. Radiation shielding and dose rate evaluation at the interim storage facility for spent fuel from Cernavoda NPP

    International Nuclear Information System (INIS)

    Stanciu, Marcela; Mateescu, Silvia; Pantazi, Doina; Penescu, Maria

    2000-01-01

    At present studies necessary to license the Interim Storage Facility for the Spent Fuel (CANDU type) from Cernavoda NPP are developed in our country.The spent fuel from Cernavoda NPP is discharged into Spent Fuel Bay in Service Building of the plant, where it remains several years for cooling. After this period, the bundles of spent fuel are to be transferred to the Interim Storage Facility.The dry interim storage solution seems to be the most appropriate variant for Cernavoda NPP.The design of the Spent Fuel Interim Storage Facility must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility achievement. In this paper we intend to present the calculation of radiation shielding at the spent fuel interim storage facility for two technical solutions: - Concrete Monolithic Module and Concrete Storage Cask. In order to quantify the fuel composition after irradiation, the isotope generation and depletion code ORIGEN 2.1 has been used, taking into account a cooling time of 7 years and 9 years, respectively, for these two variants. The shielding calculations have been performed using the computer codes QAD-5K and MICROSHIELD-4. The evaluations refer only to gamma radiation because the resulting neutron source (from (α,n) reactions and spontaneous fission) is insignificant as compared to the gamma source. The final results consist in the minimum thickness of the shielding and the corresponding external dose rates, ensuring a design average dose rate based on national and international regulations. (authors)

  2. Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities.

    Science.gov (United States)

    Glasgow, Glenn P

    2006-09-01

    Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities NCRP Report No. 151, 2005, 246 pp. (Hardcover $100). National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 400, Bethesda, MD 20814-3095. ISBN-10 0-0929600-87-8; http://www.NCRPonline.org. © 2006 American Association of Physicists in Medicine.

  3. 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)

  4. Shielded analytical laboratory activities supporting waste isolation programs

    International Nuclear Information System (INIS)

    McCown, J.J.

    1985-08-01

    The Shielded Analytical Laboratory (SAL) is a six cell manipulator-equipped facility which was built in 1962 as an addition to the 325 Radiochemistry Bldg. in the 300 Area at Hanford. The facility provides the capability for handling a wide variety of radioactive materials and performing chemical dissolutions, separations and analyses on nuclear fuels, components, waste forms and materials from R and D programs

  5. 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

  6. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1991-02-01

    The Hot Cell facility at Risoe has been in active use since 1964. During the years several types of nuclear fuels have been handled and examined: test reactor fuel pins from the Danish reactor DR3, the Norwegian Halden reactor, etc; power reactor fuel pins from several foreign reactors, including plutonium enriched pins; HTGR fuel from the Dragon reactor. All kinds of physical and chemical non-destructive and destructive post irradiation examinations have been performed. Besides, different radiotherapy sources have been produced, mainly cobalt sources. The general object of the decommissioning programme for the Hot Cell facility was to obtain a safe condition for the total building that does not require the special safety provisions. The hot cell building will be usable for other purposes after decommissioning. The facilicy comprised six concrete cells, lead cells, glove boxes, a shielded unit for temporary storage of waste, frogman area, decontamination areas, workshops, various installations of importance for safe operation of the plant, offices, etc. The tasks comprised e.g. removal of all irradiated fuel items, removal of other radioactive items, removal of contaminated equipment, and decontamination of all the cells and rooms. The goal was to decontaminate all the concrete cells to a degree where no loose contamination exists in the cells, and where the radiation level is so low, that total removal of the cell structures can be done at any time in the future without significant dose commitments. (AB)

  7. Use of a CO2 pellet non-destructive cleaning system to decontaminate radiological waste and equipment in shielded hot cells at the Bettis Atomic Power Laboratory

    International Nuclear Information System (INIS)

    Bench, T.R.

    1997-01-01

    This paper details how the Bettis Atomic Power Laboratory modified and utilized a commercially available, solid carbon dioxide (CO 2 ) pellet, non-destructive cleaning system to support the disposition and disposal of radioactive waste from shielded hot cells. Some waste materials and equipment accumulated in the shielded hot cells cannot be disposed directly because they are contaminated with transuranic materials (elements with atomic numbers greater than that of uranium) above waste disposal site regulatory limits. A commercially available CO 2 pellet non-destructive cleaning system was extensively modified for remote operation inside a shielded hot cell to remove the transuranic contaminants from the waste and equipment without generating any secondary waste in the process. The removed transuranic contaminants are simultaneously captured, consolidated, and retained for later disposal at a transuranic waste facility

  8. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  9. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  10. Computer control of shielded cell operations

    International Nuclear Information System (INIS)

    Jeffords, W.R. III.

    1987-01-01

    This paper describes in detail a computer system to remotely control shielded cell operations. System hardware, software, and design criteria are discussed. We have designed a computer-controlled buret that provides a tenfold improvement over the buret currently in service. A computer also automatically controls cell analyses, calibrations, and maintenance. This system improves conditions for the operators by providing a safer, more efficient working environment and is expandable for future growth and development

  11. Design report for shielded glove box

    International Nuclear Information System (INIS)

    Ku, J. H.; Lee, J. C.; Seo, K. S.; Bang, K. S.; Lee, D. W.; Kim, J. H.; Min, D. K.; Park, S. W.

    1999-05-01

    For the examination of spent fuels and high radioactive specimens using a specially equipped scanning electron microscope, a shielded glove box was designed and constructed at PIE facility of KAERI. This glove box consisted of shielding walls, containment box, lead glasses, manipulators, gloves, ventilation systems, doors, hot-cell specimen cask adapter, etc. It was emphasized that both the easy operation and radiation safety are important factors in the shielded glove box were installed also considered as a important factor to build the basic concept of the assembling. Two sliding doors and one hinge-type door were installed for the easy installation, operation and maintenance of scanning electron microscope. Containment box which confines the radioactive material into the box consisted of reinforced transparent glasses, aluminum frames and stainless steel plate liner. Therefore everything beyond the containment box can be seen through the lead glass which installed at the front shielding wall. All shielding walls and doors were introduced separately into the room and assembled by bolting. (author). 3 refs., 5 tabs., 18 figs

  12. 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)

  13. Seismic analysis of the mirror fusion test facility shielding vault

    International Nuclear Information System (INIS)

    Gabrielsen, B.L.; Tsai, K.

    1981-04-01

    This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

  14. 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.)

  15. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Science.gov (United States)

    Genard, Gilles; Portal, Romain; Bouchat, Virginie; Vanderperre, Serge

    2017-09-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  16. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Genard Gilles

    2017-01-01

    Full Text Available In the framework of the design studies for a new Emergency Operation Facility (EOF for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  17. Analytical study plan: Shielded Cells batch 1 campaign; Revision 1

    International Nuclear Information System (INIS)

    Bibler, N.E.; Ha, B.C.; Hay, M.S.; Ferrara, D.M.; Andrews, M.K.

    1993-01-01

    Radioactive operations in the Defense Waste Processing Facility (DWPF) will require that the Savannah River Technology Center (SRTC) perform analyses and special studies with actual Savannah River Site (SRS) high-level waste sludge. SRS Tank 42 and Tank 51 will comprise the first batch of sludge to be processed in the DWPF. Approximately 25 liters of sludge from each of these tanks will be characterized and processed in the Shielded Cells of SRTC. During the campaign, processes will include sludge characterization, sludge washing, rheology determination, mixing, hydrogen evolution, feed preparation, and vitrification of the waste. To complete the campaign, the glass will be characterized to determine its durability and crystallinity. This document describes the types of samples that will be produced, the sampling schedule and analyses required, and the methods for sample and analytical control

  18. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility

    International Nuclear Information System (INIS)

    Wind, Michael; Beck, Peter; Latocha, Marcin

    2015-01-01

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  19. Shield design of concrete wall between decay tank room and primary pump room in TRIGA facility

    International Nuclear Information System (INIS)

    Khan, M. J. H.; Rahman, M.; Haque, A.; Zulquarnain, A.; Ahmed, F. U.; Bhuiyan, S. I.

    2007-01-01

    The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II research reactor facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit (10 μSv/hr). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete

  20. Highly Efficient and Reliable Transparent Electromagnetic Interference Shielding Film.

    Science.gov (United States)

    Jia, Li-Chuan; Yan, Ding-Xiang; Liu, Xiaofeng; Ma, Rujun; Wu, Hong-Yuan; Li, Zhong-Ming

    2018-04-11

    Electromagnetic protection in optoelectronic instruments such as optical windows and electronic displays is challenging because of the essential requirements of a high optical transmittance and an electromagnetic interference (EMI) shielding effectiveness (SE). Herein, we demonstrate the creation of an efficient transparent EMI shielding film that is composed of calcium alginate (CA), silver nanowires (AgNWs), and polyurethane (PU), via a facile and low-cost Mayer-rod coating method. The CA/AgNW/PU film with a high optical transmittance of 92% achieves an EMI SE of 20.7 dB, which meets the requirements for commercial shielding applications. A superior EMI SE of 31.3 dB could be achieved, whereas the transparent film still maintains a transmittance of 81%. The integrated efficient EMI SE and high transmittance are superior to those of most previously reported transparent EMI shielding materials. Moreover, our transparent films exhibit a highly reliable shielding ability in a complex service environment, with 98 and 96% EMI SE retentions even after 30 min of ultrasound treatment and 5000 bending cycles (1.5 mm radius), respectively. The comprehensive performance that is associated with the facile fabrication strategy imparts the CA/AgNW/PU film with great potential as an optimized EMI shielding material in emerging optoelectronic devices, such as flexible solar cells, displays, and touch panels.

  1. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    OpenAIRE

    Genard Gilles; Portal Romain; Bouchat Virginie; Vanderperre Serge

    2017-01-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assump...

  2. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT

    International Nuclear Information System (INIS)

    Evans, J.F.; Blue, T.E.

    1996-01-01

    Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions open-quotes How much?close quotes and open-quotes What kind?close quotes of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room, patient open-quotes scatterer,close quotes and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h -1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel. 20 refs., 8 figs., 2 tabs

  3. SHIELD 1.0: development of a shielding calculator program in diagnostic radiology; SHIELD 1.0: desenvolvimento de um programa de calculo de blindagem em radiodiagnostico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da [Hospital Sao Lucas (PUCRS), Porto Alegre, RS (Brazil); Friedrich, Barbara Q.; Silva, Ana Maria Marques da, E-mail: ana.marques@pucrs.br [Pontificia Universidade Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil)

    2013-08-15

    In shielding calculation of radiological facilities, several parameters are required, such as occupancy, use factor, number of patients, source-barrier distance, area type (controlled and uncontrolled), radiation (primary or secondary) and material used in the barrier. The shielding design optimization requires a review of several options about the physical facility design and, mainly, the achievement of the best cost-benefit relationship for the shielding material. To facilitate the development of this kind of design, a program to calculate the shielding in diagnostic radiology was implemented, based on data and limits established by National Council on Radiation Protection and Measurements (NCRP) 147 and SVS-MS 453/98. The program was developed in C⌗ language, and presents a graphical interface for user data input and reporting capabilities. The module initially implemented, called SHIELD 1.0, refers to calculating barriers for conventional X-ray rooms. The program validation was performed by the comparison with the results of examples of shielding calculations presented in NCRP 147.

  4. Biological shielding design and qualification of concreting process for construction of electron beam irradiation facility

    International Nuclear Information System (INIS)

    Petwal, V.C.; Kumar, P.; Suresh, N.; Parchani, G.; Dwivedi, J.; Thakurta, A.C.

    2011-01-01

    A technology demonstration facility for irradiation of food and agricultural products is being set-up by RRCAT at Indore. The facility design is based on linear electron accelerator with maximum beam power of 10 kW and can be operated either in electron mode at 10 MeV or photon modes at 5/7.5 MeV. Biological shielding has been designed in accordance with NCRP 51 to achieve dose rate at all accessible points outside the irradiation vault less than the permissible limit of 0.1 mR/hr. In addition to radiation attenuation property, concrete must have satisfactory mechanical properties to meet the structural requirements. There are number of site specific variables which affect the structural, thermal and radiological properties of concrete, leading to considerable difference in actual values and design values. Hence it is essential to establish a suitable site and environmental specific process to cast the concrete and qualify the process by experimental measurement. For process qualification we have cast concrete test blocks of different thicknesses up to 3.25 m and evaluated the radiological and mechanical properties by radiometry, ultrasonic and mechanical tests. In this paper we describe the biological shielding design of the facility and analyse the results of tests carried out for qualification of the process. (author)

  5. Analytical methods and laboratory facility for the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Coleman, C.J.; Dewberry, R.A.; Lethco, A.J.; Denard, C.D.

    1985-01-01

    This paper describes the analytical methods, instruments, and laboratory that will support vitrification of defense waste. The Defense Waste Processing Facility (DWPF) is now being constructed at Savannah River Plant (SRP). Beginning in 1989, SRP high-level defense waste will be immobilized in borosilicate glass for disposal in a federal repository. The DWPF will contain an analytical laboratory for performing process control analyses. Additional analyses will be performed for process history and process diagnostics. The DWPF analytical facility will consist of a large shielded sampling cell, three shielded analytical cells, a laboratory for instrumental analysis and chemical separations, and a counting room. Special instrumentation is being designed for use in the analytical cells, including microwave drying/dissolution apparatus, and remote pipetting devices. The instrumentation laboratory will contain inductively coupled plasma, atomic absorption, Moessbauer spectrometers, a carbon analyzer, and ion chromatography equipment. Counting equipment will include intrinsic germanium detectors, scintillation counters, Phoswich alpha, beta, gamma detectors, and a low-energy photon detector

  6. A method for assessing the structural shielding in diagnostic x-ray facilities

    International Nuclear Information System (INIS)

    Almeida, F.; Martinez de la Fuente, O.; Perez, C.

    1992-01-01

    The design of each X-Ray medical facility involves, in order to guarantee the optimun levels of Radiologic safety for everybody who could be exposed during the performance of the examinations the need of assessing the required shieldings for the room which contains the emiter tubes. In such sense, this paper gives a number of criteria to calculate the structural requirements for the diverse configurations which exist in Health Centres using the method proposed by NCRP in its Report 49 as a reference. (author)

  7. Safety analysis report for the National Low-Temperature Neutron Irradiation Facility (NLTNIF) at the ORNL Bulk Shielding Reactor (BSR)

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Kerchner, H.R.; Klabunde, C.E.; Richardson, S.A.

    1986-06-01

    This report provides information concerning: the experiment facility; experiment assembly; instrumentation and controls; materials; radioactivity; shielding; thermodynamics; estimated or measured reactivity effects; procedures; hazards; and quality assurance

  8. Development of a computer code for shielding calculation in X-ray facilities

    International Nuclear Information System (INIS)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

    2014-01-01

    The construction of an effective barrier against the interaction of ionizing radiation present in X-ray rooms requires consideration of many variables. The methodology used for specifying the thickness of primary and secondary shielding of an traditional X-ray room considers the following factors: factor of use, occupational factor, distance between the source and the wall, workload, Kerma in the air and distance between the patient and the receptor. With these data it was possible the development of a computer program in order to identify and use variables in functions obtained through graphics regressions offered by NCRP Report-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) for the calculation of shielding of the room walls as well as the wall of the darkroom and adjacent areas. With the built methodology, a program validation is done through comparing results with a base case provided by that report. The thickness of the obtained values comprise various materials such as steel, wood and concrete. After validation is made an application in a real case of radiographic room. His visual construction is done with the help of software used in modeling of indoor and outdoor. The construction of barriers for calculating program resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in September / 2011

  9. A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

    OpenAIRE

    Hoff, Gabriela; Costa, Paulo Roberto

    2013-01-01

    The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and com...

  10. Removal site evaluation report on the Tower Shielding Facility at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-09-01

    This removal site evaluation report for the Tower Shielding Facility (TSF) at Oak Ridge National Laboratory was prepared to provide the Environmental Restoration Program with information necessary to evaluate whether hazardous and/or radiological contaminants in and around the Tower Shielding Facility pose a substantial risk to human health or the environment (i.e., a high probability of adverse effects) and if remedial site evaluations or removal actions are, therefore, required. The scope of the project included a review of historical evidence regarding operations and use of the facility; interviews with facility personnel concerning current and past operating practices; a site inspection; and identification of hazard areas requiring maintenance, removal, or remedial actions. Based an the findings of this removal site evaluation, adequate efforts are currently being made at the TSF to contain and control existing contamination and hazardous substances on site in order to protect human health and the environment No conditions requiring maintenance or removal actions to mitigate imminent or potential threats to human health and the environment were identified during this evaluation. Given the current conditions and status of the buildings associated with the TSF, this removal site evaluation is considered complete and terminated according to the requirements for removal site evaluation termination

  11. Neutron shield analysis and design for the PDX fusion facility

    International Nuclear Information System (INIS)

    Grimesey, R.A.; Nigg, D.W.; Scott, A.J.; Wheeler, F.J.; Jassby, D.L.; Perry, E.D.

    1979-01-01

    The basic component of the biological shield for PDX is an existing 81 cm thick high-density concrete shielding wall surrounding the machine. The principal additional shielding requirement is a roof shield over the machine to reduce air-scattered skyshine dose into the PDX control room and to the site boundary. The roof shield is designed in removable sections on a steel support structure permitting overhead crane access to major PDX components. After analysis of a number of alternate concepts, a roof shield consisting of 50 cm of water in polyethylene tanks was selected to meet design objectives of effectiveness, weight, removability, and cost

  12. Technology development for radiation shielding analysis

    International Nuclear Information System (INIS)

    Ha, Jung Woo; Lee, Jae Kee; Kim, Jong Kyung

    1986-12-01

    Radiation shielding analysis in nuclear engineering fields is an important technology which is needed for the calculation of reactor shielding as well as radiation related safety problems in nuclear facilities. Moreover, the design technology required in high level radioactive waste management and disposal facilities is faced on serious problems with rapidly glowing nuclear industry development, and more advanced technology has to be developed for tomorrow. The main purpose of this study is therefore to build up the self supporting ability of technology development for the radiation shielding analysis in order to achieve successive development of nuclear industry. It is concluded that basic shielding calculations are possible to handle and analyze by using our current technology, but more advanced technology is still needed and has to be learned for the degree of accuracy in two-dimensional shielding calculation. (Author)

  13. The clinical testing of male gonad shields. Technical report

    International Nuclear Information System (INIS)

    Church, W.W.; Burnett, B.M.

    1975-11-01

    Two types of male gonad shields, designed for use with support garments, were tested in a number of hospitals and clinics throughout the United States. The clinical evaluation consisted of: (1) measuring dose reduction with thermoluminescent dosimeters; and (2) determining acceptability of the shields for routine use in x-ray facilities, through the use of survey forms completed by patients, technologists, and facilities. The shields proved to provide a basis for a very satisfactory male gonad shield program

  14. The clinical testing of male gonad shields. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Church, W.W.; Burnett, B.M.

    1975-11-01

    Two types of male gonad shields, designed for use with support garments, were tested in a number of hospitals and clinics throughout the United States. The clinical evaluation consisted of: (1) measuring dose reduction with thermoluminescent dosimeters; and (2) determining acceptability of the shields for routine use in x-ray facilities, through the use of survey forms completed by patients, technologists, and facilities. The shields proved to provide a basis for a very satisfactory male gonad shield program. (GRA)

  15. Development of a computational code for calculations of shielding in dental facilities

    International Nuclear Information System (INIS)

    Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

    2014-01-01

    This paper is prepared in order to address calculations of shielding to minimize the interaction of patients with ionizing radiation and / or personnel. The work includes the use of protection report Radiation in Dental Medicine (NCRP-145 or Radiation Protection in Dentistry), which establishes calculations and standards to be adopted to ensure safety to those who may be exposed to ionizing radiation in dental facilities, according to the dose limits established by CNEN-NN-3.1 standard published in September / 2011. The methodology comprises the use of computer language for processing data provided by that report, and a commercial application used for creating residential projects and decoration. The FORTRAN language was adopted as a method for application to a real case. The result is a programming capable of returning data related to the thickness of material, such as steel, lead, wood, glass, plaster, acrylic, acrylic and leaded glass, which can be used for effective shielding against single or continuous pulse beams. Several variables are used to calculate the thickness of the shield, as: number of films used in the week, film load, use factor, occupational factor, distance between the wall and the source, transmission factor, workload, area definition, beam intensity, intraoral and panoramic exam. Before the application of the methodology is made a validation of results with examples provided by NCRP-145. The calculations redone from the examples provide answers consistent with the report

  16. Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

    International Nuclear Information System (INIS)

    Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

    2012-01-01

    Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

  17. Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities.

    Science.gov (United States)

    Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

    2012-10-01

    A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

  18. 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

  19. The shielding calculation for the CN guide shielding assembly in HANARO

    International Nuclear Information System (INIS)

    Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

    2006-01-01

    The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

  20. 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

  1. Studsvik thermal neutron facility

    International Nuclear Information System (INIS)

    Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

    1992-01-01

    The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

  2. Field maintenance of radiation-shielding windows at HFEF

    International Nuclear Information System (INIS)

    Tobias, D.A.

    1983-01-01

    The achievement of excellent viewing through hot-cell shielding windows does not occur by chance. Instead, it requires a well planned and executed program of field maintenance. The lack of such a program is a major factor when a hot-cell facility has poor window viewing. At HFEF, all preventive maintenance is performed by one group of trained technical-support personnel under the immediate direction of a Systems Engineer, who has responsibility for the shielding windows. Window maintenance is prescheduled and recorded by being incorporated into the computerized Maintenance Data System (MDS). Measurements of window light transmission are scheduled annually to determine glass browning or oil cloudiness conditions within the window tank. The tank oil is sampled and chemically analyzed annually to determine the moisture content, the acidity, and the probable deterioration rate caused by irradiation

  3. 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)

  4. Shielding of Medical Radiation Facilities - National Council on Radiation Protection and Measurements Reports No. 147 and No. 151

    International Nuclear Information System (INIS)

    KASE, K.R.

    2008-01-01

    The National Council on Radiation Protection and Measurements of the United States (NCRP) has issued two reports in the past 18 months that provide methods and data for designing shielding for diagnostic radiological imaging and radiation therapy facilities. These reports update previous publications on this subject with revised methods that take into account new technologies, results from measurements and new data that have been published in the last 30 years. This paper gives a brief summary of the contents of these reports, the methods recommended for determining the shielding required and the data provided to aid in the calculations

  5. Performance and Facility Background Pressure Characterization Tests of NASAs 12.5-kW Hall Effect Rocket with Magnetic Shielding Thruster

    Science.gov (United States)

    Kamhawi, Hani; Huang, Wensheng; Haag, Thomas; Shastry, Rohit; Thomas, Robert; Yim, John; Herman, Daniel; Williams, George; Myers, James; Hofer, Richard; hide

    2015-01-01

    NASA's Space Technology Mission Directorate (STMD) Solar Electric Propulsion Technology Demonstration Mission (SEP/TDM) project is funding the development of a 12.5-kW Hall thruster system to support future NASA missions. The thruster designated Hall Effect Rocket with Magnetic Shielding (HERMeS) is a 12.5-kW Hall thruster with magnetic shielding incorporating a centrally mounted cathode. HERMeS was designed and modeled by a NASA GRC and JPL team and was fabricated and tested in vacuum facility 5 (VF5) at NASA GRC. Tests at NASA GRC were performed with the Technology Development Unit 1 (TDU1) thruster. TDU1's magnetic shielding topology was confirmed by measurement of anode potential and low electron temperature along the discharge chamber walls. Thermal characterization tests indicated that during full power thruster operation at peak magnetic field strength, the various thruster component temperatures were below prescribed maximum allowable limits. Performance characterization tests demonstrated the thruster's wide throttling range and found that the thruster can achieve a peak thruster efficiency of 63% at 12.5 kW 500 V and can attain a specific impulse of 3,000 s at 12.5 kW and a discharge voltage of 800 V. Facility background pressure variation tests revealed that the performance, operational characteristics, and magnetic shielding effectiveness of the TDU1 design were mostly insensitive to increases in background pressure.

  6. Radiation field characterization and shielding studies for the ELI Beamlines facility

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, A., E-mail: a.ferrari@hzdr.de [Institute of Radiation Physics, Helmholtz-Zentrum Dresden-Rossendorf, PF 510119, 01314 Dresden (Germany); Amato, E. [Department of Radiological Sciences, Messina University (Italy); Margarone, D. [ELI Beamlines Project, Institute of Physics of the ASCR, Na Slovance 2, 18221 Prague (Czech Republic); PALS Centre, Za Slovankou, 18200 Prague (Czech Republic); Cowan, T. [Institute of Radiation Physics, Helmholtz-Zentrum Dresden-Rossendorf, PF 510119, 01314 Dresden (Germany); Korn, G. [ELI Beamlines Project, Institute of Physics of the ASCR, Na Slovance 2, 18221 Prague (Czech Republic)

    2013-05-01

    The ELI (Extreme Light Infrastructure) Beamlines facility in the Czech Republic, which is planned to complete the installation in 2015, is one of the four pillars of the ELI European project. Several laser beamlines with ultrahigh intensities and ultrashort pulses are foreseen, offering versatile radiation sources in an unprecedented energy range: laser-driven particle beams are expected to range between 1 and 50 GeV for electrons and from 100 MeV up to 3 GeV for protons. The number of particles delivered per laser shot is estimated to be 10{sup 9}–10{sup 10} for the electron beams and 10{sup 10}–10{sup 12} for the proton beams. The high energy and current values of the produced particles, together with the potentiality to operate at 10 Hz laser repetition rate, require an accurate study of the primary and secondary radiation fields to optimize appropriate shielding solutions: this is a key issue to minimize prompt and residual doses in order to protect the personnel, reduce the radiation damage of electronic devices and avoid strong limitations in the operational time. A general shielding study for the 10 PW (0.016 Hz) and 2 PW (10 Hz) laser beamlines is presented here. Starting from analytical calculations, as well as from dedicated simulations, the main electron and proton fields produced in the laser-matter interaction have been described and used to characterize the “source terms” in full simulations with the Monte Carlo code FLUKA. The secondary radiation fields have been then analyzed to assess a proper shielding. The results of this study and the proposed solutions for the beam dumps of the high energy beamlines, together with a cross-check analysis performed with the Monte Carlo code GEANT4, are presented.

  7. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

    Olson, R.E.; Thumim, A.D.

    1975-01-01

    A scintillation counter, particularly for counting gamma ray photons, includes a massive lead radiation shield surrounding a sample-receiving zone. The shield is disassembleable into a plurality of segments to allow facile installation and removal of a photomultiplier tube assembly, the segments being so constructed as to prevent straight-line access of external radiation through the shield into radiation-responsive areas. Provisions are made for accurately aligning the photomultiplier tube with respect to one or more sample-transmitting bores extending through the shield to the sample receiving zone. A sample elevator, used in transporting samples into the zone, is designed to provide a maximum gamma-receiving aspect to maximize the gamma detecting efficiency. (U.S.)

  8. Nutrient Shielding in Clusters of Cells

    Science.gov (United States)

    Lavrentovich, Maxim O.; Koschwanez, John H.; Nelson, David R.

    2014-01-01

    Cellular nutrient consumption is influenced by both the nutrient uptake kinetics of an individual cell and the cells’ spatial arrangement. Large cell clusters or colonies have inhibited growth at the cluster's center due to the shielding of nutrients by the cells closer to the surface. We develop an effective medium theory that predicts a thickness ℓ of the outer shell of cells in the cluster that receives enough nutrient to grow. The cells are treated as partially absorbing identical spherical nutrient sinks, and we identify a dimensionless parameter ν that characterizes the absorption strength of each cell. The parameter ν can vary over many orders of magnitude between different cell types, ranging from bacteria and yeast to human tissue. The thickness ℓ decreases with increasing ν, increasing cell volume fraction ϕ, and decreasing ambient nutrient concentration ψ∞. The theoretical results are compared with numerical simulations and experiments. In the latter studies, colonies of budding yeast, Saccharomyces cerevisiae, are grown on glucose media and imaged under a confocal microscope. We measure the growth inside the colonies via a fluorescent protein reporter and compare the experimental and theoretical results for the thickness ℓ. PMID:23848711

  9. 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

  10. 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

  11. Seismic proof test of shielding block walls

    International Nuclear Information System (INIS)

    Ohte, Yukio; Watanabe, Takahide; Watanabe, Hiroyuki; Maruyama, Kazuhide

    1989-01-01

    Most of the shielding block walls used for building nuclear facilities are built by dry process. When a nuclear facility is designed, seismic waves specific at each site are set as input seismic motions and they are adopted in the design. Therefore, it is necessary to assure safety of the shielding block walls for earthquake by performing anti-seismic experiments under the conditions at each site. In order to establish the normal form that can be applied to various seismic conditions in various areas, Shimizu Corp. made an actual-size test samples for the shielding block wall and confirmed the safety for earthquake and validity of normalization. (author)

  12. Assessment of the integrity of structural shielding of four computed tomography facilities in the greater Accra region of Ghana

    International Nuclear Information System (INIS)

    Nkansah, A.; Schandorf, C.; Boadu, M.; Fletcher, J. J.

    2013-01-01

    The structural shielding thicknesses of the walls of four computed tomography (CT) facilities in Ghana were re-evaluated to verify the shielding integrity using the new shielding design methods recommended by the National Council on Radiological Protection and Measurements (NCRP). The shielding thickness obtained ranged from 120 to 155 mm using default DLP values proposed by the European Commission and 110 to 168 mm using derived DLP values from the four CT manufacturers. These values are within the accepted standard concrete wall thickness ranging from 102 to 152 mm prescribed by the NCRP. The ultrasonic pulse testing of all walls indicated that these are of good quality and free of voids since pulse velocities estimated were within the range of 3.496±0.005 km s -1 . An average dose equivalent rate estimated for supervised areas is 3.4±0.27 μSv week -1 and that for the controlled area is 18.0±0.15 μSv week -1 , which are within acceptable values. (authors)

  13. Evaluation of RayXpert® for shielding design of medical facilities

    Science.gov (United States)

    Derreumaux, Sylvie; Vecchiola, Sophie; Geoffray, Thomas; Etard, Cécile

    2017-09-01

    In a context of growing demands for expert evaluation concerning medical, industrial and research facilities, the French Institute for radiation protection and nuclear safety (IRSN) considered necessary to acquire new software for efficient dimensioning calculations. The selected software is RayXpert®. Before using this software in routine, exposure and transmission calculations for some basic configurations were validated. The validation was performed by the calculation of gamma dose constants and tenth value layers (TVL) for usual shielding materials and for radioisotopes most used in therapy (Ir-192, Co-60 and I-131). Calculated values were compared with results obtained using MCNPX as a reference code and with published values. The impact of different calculation parameters, such as the source emission rays considered for calculation and the use of biasing techniques, was evaluated.

  14. Technical experience of the Belgian regulatory body regarding shielded cells

    International Nuclear Information System (INIS)

    Cortenbosch, G.; Degreef, G.; Noterman, N.; Mommaert, C.; Smidts, O.; Hasendonck, M. van; Drymael, H.; Schmitz, F.; Kennes, C.; Carlier, P.

    2010-01-01

    Radiation protection and nuclear safety are of paramount importance when operating shielded cells in hot labs and using remote handling tools. Therefore, various aspects should be ensured throughout the lifecycle of nuclear installations disposing of such type of equipment. The presented aspects are the results of the regulatory body experience with this kind of installations (licensing, operations, operating experience, events). This paper will deal with safety considerations such as filtration, fire protection, shielding, ventilation, waste management, separation of the product fluxes, leak tightness, introduction of equipment, interlocks, failsafe components, piloting software, monitoring and education/training. (orig.)

  15. Preliminary definition of the remote handling system for the current IFMIF Test Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Queral, V., E-mail: vicentemanuel.queral@ciemat.es [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Urbon, J. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Garcia, A.; Cuarental, I.; Mota, F. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Micciche, G. [CR ENEA Brasimone, I-40035 Camugnano (BO) (Italy); Ibarra, A. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Rapisarda, D. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Casal, N. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain)

    2011-10-15

    A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

  16. Preliminary definition of the remote handling system for the current IFMIF Test Facilities

    International Nuclear Information System (INIS)

    Queral, V.; Urbon, J.; Garcia, A.; Cuarental, I.; Mota, F.; Micciche, G.; Ibarra, A.; Rapisarda, D.; Casal, N.

    2011-01-01

    A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

  17. Design of radiation shielding for the proton therapy facility at the National Cancer Center in Korea

    International Nuclear Information System (INIS)

    Kim, J. W.; Kwon, J. W.; Lee, J.

    2005-01-01

    The design of radiation shielding was evaluated for a proton therapy facility being established at the National Cancer Center in Korea. The proton beam energy from a 230 MeV cyclotron is varied for therapy using a graphite target. This energy variation process produces high radiation and thus thick shielding walls surround the region. The evaluation was first carried out using analytical expressions at selected locations. Further detailed evaluations have been performed using the Monte Carlo method. Dose equivalent values were calculated to be compared with analytical results. The analytical method generally yielded more conservative values. With consideration of adequate occupancy factors annual dose equivalent rates are kept -1 in all areas. Construction of the building is expected to be completed near the end of 2004 and the installation of therapy equipments will begin a few months later. (authors)

  18. Shielded scanning electron microscope for radioactive samples

    International Nuclear Information System (INIS)

    Crouse, R.S.; Parsley, W.B.

    1977-01-01

    A small commercial SEM had been successfully shielded for examining radioactive materials transferred directly from a remote handling facility. Relatively minor mechanical modifications were required to achieve excellent operation. Two inches of steel provide adequate shielding for most samples encountered. However, samples reading 75 rad/hr γ have been examined by adding extra shielding in the form of tungsten sample holders and external lead shadow shields. Some degradation of secondary electron imaging was seen but was adequately compensated for by changing operating conditions

  19. Satif-3: Shielding aspects of accelerators, targets, and irradiation facilities. Tohoku University, Sendai, Japan, 12--13 May 1997

    International Nuclear Information System (INIS)

    1998-01-01

    Particle accelerators have evolved over the last 50 years from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today, they cover a wide range of applications, from television and computer displays in households to the investigation of the origin and structure of matter. It has become common practice to use them for material science and medical applications. In recent years, requirements from new technological and research applications have emerged, giving rise to new radiation shielding aspects and problems. These proceedings review recent progress in radiation shielding of accelerator facilities, evaluate advancements and discuss further developments needed with respect to international co-operation in this field

  20. Radiation protection/shield design

    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. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion 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 will be discussed. (orig.) [de

  1. Remote operation and maintenance demonstration facility at ORNL

    International Nuclear Information System (INIS)

    Harvey, H.W.; Floyd, S.D.; Kuban, D.P.; Singletary, B.H.; Stradley, J.G.

    1978-01-01

    The Remote Operation and Maintenance Facility is a versatile facility arranged to mock up various hot cell configurations. Modular units of simulated shielding and viewing windows were built to provide flexibility in arrangement. The facility is fully equipped with hoists, manipulators, television, and other basic equipment and services necessary to provide capability for both remote operation and maintenance of several selected functional process equipment groups

  2. 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.

  3. 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)

  4. Facility effluent monitoring plan for the 325 Facility

    International Nuclear Information System (INIS)

    1998-01-01

    The Applied Chemistry Laboratory (325 Facility) houses radiochemistry research, radioanalytical service, radiochemical process development, and hazardous and mixed hazardous waste treatment activities. The laboratories and specialized facilities enable work ranging from that with nonradioactive materials to work with picogram to kilogram quantities of fissionable materials and up to megacurie quantities of other radionuclides. The special facilities include two shielded hot-cell areas that provide for process development or analytical chemistry work with highly radioactive materials, and a waste treatment facility for processing hazardous, mixed, low-level, and transuranic wastes generated by Pacific Northwest Laboratory. Radioactive material storage and usage occur throughout the facility and include a large number of isotopes. This material is in several forms, including solid, liquid, particulate, and gas. Some of these materials are also heated during testing which can produce vapors. The research activities have been assigned to the following activity designations: High-Level Hot Cell, Hazardous Waste Treatment Unit, Waste Form Development, Special Testing Projects, Chemical Process Development, Analytical Hot Cell, and Analytical Chemistry. The following summarizes the airborne and liquid effluents and the results of the Facility Effluent Monitoring Plan (FEMP) determination for the facility. The complete monitoring plan includes characterization of effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements

  5. Cerium oxide nanoparticles, combining antioxidant and UV shielding properties, prevent UV-induced cell damage and mutagenesis

    Science.gov (United States)

    Caputo, Fanny; de Nicola, Milena; Sienkiewicz, Andrzej; Giovanetti, Anna; Bejarano, Ignacio; Licoccia, Silvia; Traversa, Enrico; Ghibelli, Lina

    2015-09-01

    Efficient inorganic UV shields, mostly based on refracting TiO2 particles, have dramatically changed the sun exposure habits. Unfortunately, health concerns have emerged from the pro-oxidant photocatalytic effect of UV-irradiated TiO2, which mediates toxic effects on cells. Therefore, improvements in cosmetic solar shield technology are a strong priority. CeO2 nanoparticles are not only UV refractors but also potent biological antioxidants due to the surface 3+/4+ valency switch, which confers anti-inflammatory, anti-ageing and therapeutic properties. Herein, UV irradiation protocols were set up, allowing selective study of the extra-shielding effects of CeO2vs. TiO2 nanoparticles on reporter cells. TiO2 irradiated with UV (especially UVA) exerted strong photocatalytic effects, superimposing their pro-oxidant, cell-damaging and mutagenic action when induced by UV, thereby worsening the UV toxicity. On the contrary, irradiated CeO2 nanoparticles, via their Ce3+/Ce4+ redox couple, exerted impressive protection on UV-treated cells, by buffering oxidation, preserving viability and proliferation, reducing DNA damage and accelerating repair; strikingly, they almost eliminated mutagenesis, thus acting as an important tool to prevent skin cancer. Interestingly, CeO2 nanoparticles also protect cells from the damage induced by irradiated TiO2, suggesting that these two particles may also complement their effects in solar lotions. CeO2 nanoparticles, which intrinsically couple UV shielding with biological and genetic protection, appear to be ideal candidates for next-generation sun shields.

  6. Remote Operation and Maintenance Demonstration Facility at ORNL

    International Nuclear Information System (INIS)

    Harvey, H.W.; Floyd, S.D; Kuban, D.P.; Singletary, B.H.; Stradley, J.G.

    1978-01-01

    The Remote Operation and Maintenance Facility is a versatile facility arranged to mock-up various hot-cell configurations. Modular units of simulated shielding and viewing windows were built to provide flexibility in arrangement. The facility is fully equipped with hoists, manipulators, television, and the other basic equipment and services necessary to provide capability for both remote operation and maintenance of several selected functional process equipment groups. 6 figures

  7. SEM/TIMS analysis trials on hotswipe samples taken from a shielded cell at Harwell

    International Nuclear Information System (INIS)

    Tushingham, J.; Vatter, I.; Cooke, R.

    1998-09-01

    The IAEA require advanced techniques and procedures for the detection of traces of actinides to be applied to their environmental sampling programme for nuclear safeguards as a means to detect undeclared activities. 'Swipe' samples taken from within nuclear facilities by IAEA inspectors require analysis to determine their actinide content and composition by bulk and particle measurements. The use of analytical equipment capable of analysing individual particles, particularly of actinides, is essential to optimise the IAEA's aim to monitor Member State's nuclear activities more proficiently. A trial has been undertaken at the Harwell Laboratory of AEA Technology to establish the efficacy of scanning electron microscopy (SEM) and thermal ionisation mass spectrometry (TIMS) for the particle and bulk characterisation, respectively, of actinides on samples taken from within a shielded cell. These measurements were supported by γ-spectrometry and α-spectrometry. 'Hotswipe' samples taken from within a shielded cell with a well-known recent history have been prepared for particle and bulk analysis. SEM has been used to characterise individual particles from the swipe samples and the results have been related to known cell activities. Samples were prepared for SEM using a simple procedure to minimise the potential for sample contamination. The method proved to be capable of identifying 1 μm particles that contained U, Pu, Pa and Np. The measurement of U/Pu ratios was limited to particles that contained >2% Pu in U by weight. TIMS, together with alpha spectrometry, has been used to determine the bulk actinide composition of the samples whilst gamma spectrometry has been used to determine the fission product composition. Further work to improve the potential of SEM, and also secondary ionisation mass spectrometry (SIMS), for the measurement of hotswipe samples has been proposed. (author)

  8. Fabrication and electromagnetic interference shielding effectiveness of polymeric composites filled with silver-coated microorganism cells

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Mingming, E-mail: lan_mingming@163.com [College of Mechanical and Electrical Engineering, Henan Agricultural University, Zhengzhou 450002 (China); Zhang, Deyuan; Cai, Jun; Hu, Yanyan; Yuan, Liming [Bionic and Micro/Nano/Bio Manufacturing Technology Research Center, School of Mechanical Engineering and Automation, Beihang University, Beijing 100191 (China)

    2014-07-01

    In this paper, helical silver-coated Spirulina cells were used as conductive fillers for the fabrication of polymeric composites. The morphology and composition of the coated Spirulina cells were analyzed with scanning electron microscope and energy dispersive X-ray spectrometer. The densities of silver-coated Spirulina cells were measured using the standard Archimedes method with distilled water. The electrical resistivity was measured by four-probe technique using ammeter and voltmeter whereas electromagnetic interference shielding effectiveness was measured by four-port method using vector network analyzer and coaxial-airline sample holder. The results showed that the silver-coated Spirulina cells with different coating thickness were lightweight fillers compared to the other typical conductive particles. The polymeric composites could achieve good conductivity at the lower content of silver-coated Spirulina cells owing to their helical shape. The shielding effectiveness of polymeric composites had a strong dependence on their conductivity. At the coating thickness of 0.96 μm and the content of 40 vol%, the shielding effectiveness could reach above 74.3 dB in entire test wave band.

  9. SP-100 reactor cell activation

    International Nuclear Information System (INIS)

    Wilcox, A.D.

    1991-09-01

    There are plans to test the SP-100 space reactor for 2 yr in the test facility shown in Figure 1. The vacuum vessel will be in the reactor experiment (RX) cell surrounded by an inert gas atmosphere. It is proposed that the reactor test cell could contain removable-water- shielding tanks to reduce the residual activation dose rates in the test cell after the tests are completed. This reduction will allow the facility to be considered for other uses after the SP-100 tests are completed. The radiation dose rates in the test cell were calculated for several configurations of water-shielding tanks to help evaluate this concept

  10. 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.)

  11. Analytical cell decontamination and shielding window refurbishment. Final report, March 1984-March 1985

    International Nuclear Information System (INIS)

    Smokowski, R.T.

    1985-01-01

    This is a report on the decontamination and refurbishment of five inactive contaminated analytical cells and six zinc bromide filled shielding windows. The analytical cells became contaminated during the nuclear fuel reprocessing carried out by Nuclear Fuel Services from 1966 to 1972. The decontamination and decommissioning (D and D) work was performed in these cells to make them useful as laboratories in support of the West Valley Demonstration Project. To accomplish this objective, unnecessary equipment was removed from these cells. Necessary equipment and the interior of each cell were decontaminated and repaired. The shielding windows, essentially tanks holding zinc bromide, were drained and disassembled. The deteriorated, opaque zinc bromide was refined to optical clarity and returned to the tanks. All wastes generated in this operation were characterized and disposed of properly. All the decontamination and refurbishment was accomplished within 13 months. The Analytical Hot Cell has been turned over to Analytical Chemistry for the performance high-level waste (HLW) characterization analysis

  12. Shield calculations, optimization vs. paradigm

    International Nuclear Information System (INIS)

    Cornejo D, N.; Hernandez S, A.; Martinez G, A.

    2006-01-01

    Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of μSv.h -1 , independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)

  13. Neutron shielding calculations in a proton therapy facility based on Monte Carlo simulations and analytical models: Criterion for selecting the method of choice

    International Nuclear Information System (INIS)

    Titt, U.; Newhauser, W. D.

    2005-01-01

    Proton therapy facilities are shielded to limit the amount of secondary radiation to which patients, occupational workers and members of the general public are exposed. The most commonly applied shielding design methods for proton therapy facilities comprise semi-empirical and analytical methods to estimate the neutron dose equivalent. This study compares the results of these methods with a detailed simulation of a proton therapy facility by using the Monte Carlo technique. A comparison of neutron dose equivalent values predicted by the various methods reveals the superior accuracy of the Monte Carlo predictions in locations where the calculations converge. However, the reliability of the overall shielding design increases if simulation results, for which solutions have not converged, e.g. owing to too few particle histories, can be excluded, and deterministic models are being used at these locations. Criteria to accept or reject Monte Carlo calculations in such complex structures are not well understood. An optimum rejection criterion would allow all converging solutions of Monte Carlo simulation to be taken into account, and reject all solutions with uncertainties larger than the design safety margins. In this study, the optimum rejection criterion of 10% was found. The mean ratio was 26, 62% of all receptor locations showed a ratio between 0.9 and 10, and 92% were between 1 and 100. (authors)

  14. Cerium oxide nanoparticles, combining antioxidant and UV shielding properties, prevent UV-induced cell damage and mutagenesis

    KAUST Repository

    Caputo, Fanny

    2015-08-20

    Efficient inorganic UV shields, mostly based on refracting TiO2 particles, have dramatically changed the sun exposure habits. Unfortunately, health concerns have emerged from the pro-oxidant photocatalytic effect of UV-irradiated TiO2, which mediates toxic effects on cells. Therefore, improvements in cosmetic solar shield technology are a strong priority. CeO2 nanoparticles are not only UV refractors but also potent biological antioxidants due to the surface 3+/4+ valency switch, which confers anti-inflammatory, anti-ageing and therapeutic properties. Herein, UV irradiation protocols were set up, allowing selective study of the extra-shielding effects of CeO2vs. TiO2 nanoparticles on reporter cells. TiO2 irradiated with UV (especially UVA) exerted strong photocatalytic effects, superimposing their pro-oxidant, cell-damaging and mutagenic action when induced by UV, thereby worsening the UV toxicity. On the contrary, irradiated CeO2 nanoparticles, via their Ce3+/Ce4+ redox couple, exerted impressive protection on UV-treated cells, by buffering oxidation, preserving viability and proliferation, reducing DNA damage and accelerating repair; strikingly, they almost eliminated mutagenesis, thus acting as an important tool to prevent skin cancer. Interestingly, CeO2 nanoparticles also protect cells from the damage induced by irradiated TiO2, suggesting that these two particles may also complement their effects in solar lotions. CeO2 nanoparticles, which intrinsically couple UV shielding with biological and genetic protection, appear to be ideal candidates for next-generation sun shields. © The Royal Society of Chemistry 2015.

  15. Measurement of concentrations of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kawasaki, Katsuya; Kikuchi, Masamitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harada, Yasunori

    1997-07-01

    Measurement has been made to study distributions of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility. Core boring was carried out at seven positions to take samples from the concrete shield, and {gamma}-ray counting rates and {gamma}-ray spectra of these samples were measured with a NaI(Tl) detector and a Ge semiconductor detector, respectively. The following radionuclides were detected in the concrete samples: {sup 60}Co, {sup 134}Cs, {sup 152}Eu and {sup 154}Eu generated through thermal neutron capture reaction, and {sup 22}Na and {sup 54}Mn generated through nuclear reactions by bremsstrahlung and fast neutrons. The relation between the distributions of {gamma}-ray emitters, as a function of the depth of concrete, and the positions of core boring is discussed. (author)

  16. Rotating shield ceiling for the compact ignition tokamak test cell

    International Nuclear Information System (INIS)

    Commander, J.C.

    1986-01-01

    For the next phase of the United States fusion program, a compact, high-field, toroidal ignition machine with liquid nitrogen cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed. The CIT machine will be housed in a test cell with design features developed during preconceptual design. Configured as a right cylinder, the selected test cell design features: a test cell and basement with thick concrete shielding walls, and floor; leak tight tritium seals; and operational characteristics well suited to the circular CIT machine configuration and radially oriented ancillary equipment and systems

  17. The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources (SHARS)

    International Nuclear Information System (INIS)

    Liebenberg, G.R.; Al-Mughrabi, M.

    2008-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell would allow source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at single sites in each IAEA Member State. The mobile hot cell and related equipment is transported in two shipping containers to a specific country where the following process takes place: 1-) Assembly of hot cell; 2-) Removal of SHARS from working shields, encapsulation into a stainless steel capsule and placement into a long term storage shield; 3-) Conditioning of any other spent sources the country may require; 4-) Dismantling of the hot cell; 5-) Shipping equipment out of country. The operation in a specific country is planned to be executed over a three week period. This presentation will discuss the development of the mobile hot cell facility as well as the demonstration of the state of readiness of the system for manipulation of SHARS and the planned execution of the conditioning operations. As a result of this project, excess SHARS could be managed safely and securely and possibly be more easily repatriated to their country of origin for appropriate final disposition. (author)

  18. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range; Sicherheitstechnische Analyse und Auslegungsaspekte von Abschirmungen gegen Teilchenstrahlung am Beispiel von Spallationsanlagen im Megawatt Bereich

    Energy Technology Data Exchange (ETDEWEB)

    Hanslik, R.

    2006-08-15

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

  19. Subsurface Shielding Source Term Specification Calculation

    International Nuclear Information System (INIS)

    S.Su

    2001-01-01

    The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations

  20. Shielding methods development in the United States

    International Nuclear Information System (INIS)

    Mynatt, F.R.

    1977-01-01

    A generalized shielding methodology has been developed in the U.S.A. that is adaptable to the shielding analyses of all reactor types. Thus far used primarily for liquid-metal fast breeder reactors, the methodology includes several component activities: (1) developing methods for calculating radiation transport through reactor-shield systems; (2) processing cross-section libraries; (3) performing design calculations for specific systems; (4) performing and analyzing pertinent integral experiments; (5) performing sensitivity studies on both the design calculations and the experimental analyses; and, finally, (6) calculating shield design parameters and their uncertainties. The criteria for the methodology are a 5 to 10 percent accuracy for responses at locations near the core and a factor of 2 accuracy for responses at distant locations. The methodology has been successfully adapted to most in-vessel and ex-vessel problems encountered in the shield analyses of the Fast Flux Test Facility and the Fast Flux Test Facility and the Clinch River Breeder Reactor; however, improved techniques are needed for calculating regions in which radiation streaming is dominant. Areas of the methodology in which significant progress has recently been made are those involving the development of cross-section libraries, sensitivity analysis methods, and transport codes

  1. Monte Carlo simulations for the shielding of the future high-intensity accelerator facility FAIR at GSI.

    Science.gov (United States)

    Radon, T; Gutermuth, F; Fehrenbacher, G

    2005-01-01

    The Gesellschaft für Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of approximately 15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam.

  2. Monte Carlo simulations for the shielding of the future high-intensity accelerator facility fair at GSI

    International Nuclear Information System (INIS)

    Radon, T.; Gutermuth, F.; Fehrenbacher, G.

    2005-01-01

    The Gesellschaft fuer Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of ∼15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam. (authors)

  3. 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''

  4. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  5. Experimental validation of thermal design of top shield for a pool type SFR

    International Nuclear Information System (INIS)

    Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

    2016-01-01

    Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

  6. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    1993-05-01

    Hot-cell shielding walls consist of building blocks made of lead according to DIN 25407 part 1, and of special elements according to DIN 25407 part 2. Alpha-gamma cells can be built using elements for protective contamination boxes according to DIN 25480 part 1. This standards document intends to provide planning engineers, manufacturers, future users and the competent authorities and experts with a basis for the design of hot cells with lead shielding walls and the design of hot-cell equipment. (orig./HP) [de

  7. Methods and procedures for shielding analyses for the SNS

    International Nuclear Information System (INIS)

    Popova, I.; Ferguson, F.; Gallmeier, F.X.; Iverson, E.; Lu, Wei

    2011-01-01

    In order to provide radiologically safe Spallation Neutron Source operation, shielding analyses are performed according to Oak Ridge National Laboratory internal regulations and to comply with the Code of Federal Regulations. An overview of on-going shielding work for the accelerator facility and neutrons beam lines, methods used for the analyses, and associated procedures and regulations are presented. Methods used to perform shielding analyses are described as well. (author)

  8. Design of a PET/CT facility considering the shielding calculation in accordance with AAPM TG-108; Diseno de una instalacion PET/CT considerando el calculo de blindaje segun AAPM TG-108

    Energy Technology Data Exchange (ETDEWEB)

    Guevara R, V. Y.; Romero C, N. [Empresa QC DOSE S. A. C., Av. Tomas Marsano 1915, Surquillo, Lima 34 (Peru); Berrocal T, M., E-mail: vguevara@qcdose.com [Universidad Nacional Mayor de San Marcos, C. German Amezaga 375, Edif. Jorge Basadre, Ciudad Universitaria, Lima 1 (Peru)

    2014-08-15

    A Positron Emission Tomography / Computed Tomography facility may require protection barriers on floor, ceiling and walls, because the patient becomes a radioactive source that emits photons of 0.511 MeV, after having received a radiopharmaceutical, usually F-18 fluorodeoxyglucose (F-18 FDG). This work has as objective to propose the design of a PET/CT facility, taking into account technical and radiation protection considerations applied internationally, and also develop the necessary shielding for such installation by applying as published by the American Association of Physicists in Medicine Task Group Report 108. A shielding spreadsheet in Excel program was developed with reference to the recommendations of the AAPM TG - 08, to determine the shielding required for the walls, floor and ceiling. For fixing the radiation levels in the shielding calculation has been considered the actual restrictions for the occupationally exposed personnel (100 μSv/week) as well as the people in general (20 μSv/ week). The radiopharmaceutical used as a reference for the shielding calculation was the F-18 FDG. With the assistance of an architectural plan were determined distances from potential sources of radiation in facility (uptake and image acquisition living rooms) to points of interest around them. Finally the thickness of the protective barriers in lead and concrete necessary to achieve the established radiation levels were calculated and these results were stored in a table. This paper shows that technical aspects considered in the design of the installation and environments distribution can improve work processes within the PET/CT facility, consequently resulting in a reduction of the dose levels for people in general. (author)

  9. The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources

    International Nuclear Information System (INIS)

    Liebenberg, G.R.; Al-Mughrabi, M.

    2010-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Administration (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused Spent High Activity Radioactive Sources (SHARS) in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell allows source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at national radioactive waste storage facilities. (authors)

  10. Radiation shielding activities at the OECD/Nuclear Energy Agency

    International Nuclear Information System (INIS)

    Sartori, Enrico; Vaz, Pedro

    2000-01-01

    The OECD Nuclear Energy Agency (NEA) has devoted considerable effort over the years to radiation shielding issues. The issues are addressed through international working groups. These activities are carried out in close co-ordination and co-operation with the Radiation Safety Information Computational Center (RSICC). The areas of work include: basic nuclear data activities in support of radiation shielding, computer codes, shipping cask shielding applications, reactor pressure vessel dosimetry, shielding experiments database. The method of work includes organising international code comparison exercises and benchmark studies. Training courses on radiation shielding computer codes are organised regularly including hands-on experience in modelling skills. The scope of the activity covers mainly reactor shields and spent fuel transportation packages, but also fusion neutronics and in particular shielding of accelerators and irradiation facilities. (author)

  11. 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.)

  12. A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

    Energy Technology Data Exchange (ETDEWEB)

    Hoff, Gabriela [Pontificia Univ. Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil). Grupo de Experimentacao e Simulacao Computacional em Fisica Medica; Costa, Paulo Roberto, E-mail: pcosta@if.usp.br [Universidade de Sao Paulo (IF/USP), SP (Brazil). Dept. de Fisica Nuclear. Lab. de Dosimetria das Radiacoes e Fisica Medica

    2013-03-15

    The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

  13. A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

    International Nuclear Information System (INIS)

    Hoff, Gabriela; Costa, Paulo Roberto

    2013-01-01

    The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

  14. Hot-cell verification facility

    International Nuclear Information System (INIS)

    Eschenbaum, R.A.

    1981-01-01

    The Hot Cell Verification Facility (HCVF) was established as the test facility for the Fuels and Materials Examination Facility (FMEF) examination equipment. HCVF provides a prototypic hot cell environment to check the equipment for functional and remote operation. It also provides actual hands-on training for future FMEF Operators. In its two years of operation, HCVF has already provided data to make significant changes in items prior to final fabrication. It will also shorten the startup time in FMEF since the examination equipment will have been debugged and operated in HCVF

  15. Neutron shielding material based on colemanite and epoxy resin

    International Nuclear Information System (INIS)

    Okuno, K.

    2005-01-01

    In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)

  16. Bulk Shielding Calculation for 90 .deg. Bending Section of RISP

    Energy Technology Data Exchange (ETDEWEB)

    Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of); Ko, S. K. [Univ. of Ulsan, Ulsan (Korea, Republic of)

    2014-10-15

    The charge state of {sup 238}U beams with maximum intensity was 79+ among multi-charge states of 70+ to 89+, which were estimated by using LISE++ code. The bending section consists of twenty four quadrupoles, two dipoles, two two-cell type superconducting RF cavities and eleven slits. The complicated radiation environment is caused by the beam losses occurred normally during the stripping process and when the produced {sup 238}U beams are transported along the beam line. Secondary radiations generated by {sup 238}U beams irradiation are very important for predicting the prompt and residual doses and the radiation damage at the component. The production characteristics of neutron and photon from thin carbon and thick iron were studied to set up the shielding strategy. The dose estimation was done to the pre-designed the tunnel structure. In these calculations, major Monte Carlo codes, PHITS and FLUKA, were used. The present study provided information of shielding analysis for the 90 .deg. bending section of RISP facility. The source term was evaluated to determine fundamental parameter of the shielding analysis using PHITS and FLUKA codes. And the distribution of the dose rate at the outside of thick shielding wall was presented.

  17. Neutron dose equivalent next to the target shield of a neutron therapy facility using an LET counter

    International Nuclear Information System (INIS)

    Stinchcomb, T.G.; Kuchnir, F.T.

    1981-01-01

    The use of a spherical tissue-equivalent proportional counter for measurements of the lineal energy (y) and derivations of the linear energy transfer (LET) for fast neutrons has the advantage of giving distributions of dose and dose equivalent as functions of either LET or y. A measurement next to the target shielding of the neutron therapy facility at the University of Chicago Hospitals and Clinics (UCHC) is described, and the data processing is outlined. The distributions are presented and compared to those from measurements in the neutron beam. The average quality factors are presented

  18. Hot cell verification facility update

    International Nuclear Information System (INIS)

    Titzler, P.A.; Moffett, S.D.; Lerch, R.E.

    1985-01-01

    The Hot Cell Verification Facility (HCVF) provides a prototypic hot cell mockup to check equipment for functional and remote operation, and provides actual hands-on training for operators. The facility arrangement is flexible and assists in solving potential problems in a nonradioactive environment. HCVF has been in operation for six years, and the facility is a part of the Hanford Engineering Development Laboratory

  19. Removal, transportation and disposal of the Millstone 2 neutron thermal shield

    International Nuclear Information System (INIS)

    Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.

    1985-01-01

    Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility

  20. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  1. The SNS target station preliminary Title I shielding analyses

    International Nuclear Information System (INIS)

    Johnson, J.O.; Santoro, R.T.; Lillie, R.A.; Barnes, J.M.; McNeilly, G.S.

    2000-01-01

    The Department of Energy (DOE) has given the Spallation Neutron Source (SNS) project approval to begin Title I design of the proposed facility to be built at Oak Ridge National Laboratory (ORNL). During the conceptual design phase of the SNS project, the target station bulk-biological shield was characterized and the activation of the major targets station components was calculated. Shielding requirements were assessed with respect to weight, space, and dose-rate constraints for operating, shut-down, and accident conditions utilizing the SNS shield design criteria, DOE Order 5480.25, and requirements specified in 10 CFR 835. Since completion of the conceptual design phase, there have been major design changes to the target station as a result of the initial shielding and activation analyses, modifications brought about due to engineering concerns, and feedback from numerous external review committees. These design changes have impacted the results of the conceptual design analyses, and consequently, have required a re-investigation of the new design. Furthermore, the conceptual design shielding analysis did not address many of the details associated with the engineering design of the target station. In this paper, some of the proposed SNS target station preliminary Title I shielding design analyses will be presented. The SNS facility (with emphasis on the target station), shielding design requirements, calculational strategy, and source terms used in the analyses will be described. Preliminary results and conclusions, along with recommendations for additional analyses, will also be presented. (author)

  2. 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.

  3. Evaluation and Verification of a Biological Shield in a SHARS Unit

    International Nuclear Information System (INIS)

    Dhlomo, S.V.; Swart, H.S.

    2008-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. This unit, designed and manufactured by the South African Nuclear Energy Corporation (Necsa), can be assembled, disassembled and packed inside two ISO containers and transported to the desired destination with relative ease. The unit was also licensed by the South African Regulator, the Department of Health (DoH), Directorate Radiation Control. This facility is used for the recovery and conditioning of orphaned/ abandoned or spent high activity radioactive sources from teletherapy units, gamma irradiators, and brachytherapy units. The hot cell was designed for a 3,7 E+13 Bq (1000 Ci) activity although it was demonstrated that it can handle activities of more than 7,4 E+13 Bq (2000 Ci) with ease. The biological shield of the SHARS facility consists of river sand sandwiched between metal plates, and a viewing window filled with a 50% zinc bromide solution. The shielding effectiveness of the river sand is evaluated experimentally by determining its density. The experimentally measured dose rates are compared to the dose rates estimated by computational codes. (authors)

  4. Evaluation and Verification of a Biological Shield in a SHARS Unit

    Energy Technology Data Exchange (ETDEWEB)

    Dhlomo, S.V.; Swart, H.S. [Compliance Management Department, Nuclear Liabilities Management, South African Nuclear Energy Corporation, P.O. Box 582, Pretoria 0001 (South Africa)

    2008-07-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. This unit, designed and manufactured by the South African Nuclear Energy Corporation (Necsa), can be assembled, disassembled and packed inside two ISO containers and transported to the desired destination with relative ease. The unit was also licensed by the South African Regulator, the Department of Health (DoH), Directorate Radiation Control. This facility is used for the recovery and conditioning of orphaned/ abandoned or spent high activity radioactive sources from teletherapy units, gamma irradiators, and brachytherapy units. The hot cell was designed for a 3,7 E+13 Bq (1000 Ci) activity although it was demonstrated that it can handle activities of more than 7,4 E+13 Bq (2000 Ci) with ease. The biological shield of the SHARS facility consists of river sand sandwiched between metal plates, and a viewing window filled with a 50% zinc bromide solution. The shielding effectiveness of the river sand is evaluated experimentally by determining its density. The experimentally measured dose rates are compared to the dose rates estimated by computational codes. (authors)

  5. 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.

  6. 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)

  7. Construction of concrete hot cells; requirements for shielding windows for concrete walls with different densities

    International Nuclear Information System (INIS)

    1987-10-01

    The shielding windows form part of the basic equipment of hot cells for remote handling, as defined in standard DIN 25 420 part 1. The draft standard in hand is intended to specify the design and manufacture requirements, especially with regard to main dimensions, sight quality, shielding effects, and radiation resistance. The standard refers to three types of shielding window with surface area design (product of density and wall thickness) corresponding to concrete walls of the densities 2.4, 3.4, and 4.0 g/cm 3 . The windows fit to three types of concrete of common usage, and the design is made for Co-60 radiation, with attenuation factors of about 10 4 , 10 6 , or 10 7 . For concrete walls with densities between these data, a shielding window suitable to the next higher density data is to be chosen. (orig./HP) [de

  8. Repository Waste Package Transporter Shielding Weight Optimization

    International Nuclear Information System (INIS)

    C.E. Sanders; Shiaw-Der Su

    2005-01-01

    The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight

  9. Shielding calculation of slow extracted beam facility at KEK proton synchrotron

    International Nuclear Information System (INIS)

    Hirabayashi, Hiromi; Katoh, Kazuaki

    1978-01-01

    The KEK proton synchrotron has two external beam lines, i.e. a fast extracted beam line for a bubble chamber and a slow extracted beam line for counter experiments. The maximum total intensity of the slow beam is estimated as 5 x 10 12 protons per sec. For beam losses along the line, shielding calculation was made, and on the basis of these results, adequacy of the current shielding construction plans was discussed. (Mori, K.)

  10. Safety analysis of DUPIC fuel development facility

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

    2001-01-01

    Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

  11. 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

  12. Shielding analysis of the IEM cell offset adapter plate

    International Nuclear Information System (INIS)

    Simons, R.L.

    1995-01-01

    The adapter plate for the Interim Examination and Maintenance (IEM) cell ten foot ceiling valve was modified so that the penetration through the valve is offset to the north side of the steel plate. The modifications required that the shielding effectiveness be evaluated for several operating conditions. The highest gamma ray dose rate (51 mrem/hr) occurs when a Core Component Container (CCC) with six high burn-up driver fuel assemblies is transferred into or out of Solid Waste Cask (SWC). The neutron dose rate at the same source location is 2.5 mrem/hr. The total dose rate during the transfer is less than the 200 mrem/hr limit. If the ten foot ceiling valve is closed, the dose rate with twelve DFA in the cell will be less than 0.1 mrem/hr. However, with the ceiling valve open the dose rate will be as high as 12 mrem/hr. The latter condition will require controlled access to the area around the offset adapter plate when the ceiling valve is open. It was found that gaps in the shield block around the SWC floor valve will allow contact dose rates as high as 350 mrem/hr during the transfer of a fully loaded CCC. Although this situation does not pertain to the offset adapter plate, it will require controlled access around the SWC valve during the transfer of a fully loaded CCC

  13. Conceptual structure design of experimental facility for advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

    2003-01-01

    A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

  14. High Density Radiation Shielding Concretes for Hot Cells of 99mTc Project

    International Nuclear Information System (INIS)

    Sakr, K.

    2006-01-01

    High density concrete [more than 3.6 ton/m 3 (3.6x10 3 kg/m 3 )] was prepared to be used as a radiation shielding concrete (RSC) for hot-cells in gel technetium project at inshas to attenuate gamma radiation emitted from radioactive sources. different types of concrete were prepared by mixing local mineral aggregates mainly gravel and ilmenite . iron shots were added to the concrete mixture proportion as partial replacement of heavy aggregates to increase its density. the physical properties of prepared concrete in both plastic and hardened phases were investigated. compressive strength and radiation attenuation of gamma rays were determined. Results showed that ilmenite concrete mixed with iron shots had the highest density suitable to be use as RSC according to the chinese hot cell design requirements. Recommendations to avoid some technical problems of manufacturing radiation shielding concrete were maintained

  15. New mass-spectrometric facility for the analysis of highly radioactive samples

    International Nuclear Information System (INIS)

    Warmack, R.J.; Landau, L.; Christie, W.H.; Carter, J.A.

    1981-01-01

    A new facility has been completed for the analysis of highly radioactive, gamma-emitting solid samples. A commercial spark-source mass spectrometer was adapted for remote handling and loading. Electrodes are prepared in a hot cell and transported to the adjacent lead-shielded source for analysis. The source was redesigned for ease of shielding, loading, and maintenance. Both solutions and residues from irradiated nuclear fuel dissolutions have been analyzed for elemental concentrations to < 1 ppM; isotopic data have also been obtained

  16. Lead shielded cells for the spectrographic analysis of radioisotope solutions

    International Nuclear Information System (INIS)

    Roca, M.; Capdevila, C.; Cruz, F. de la

    1967-01-01

    Two lead shielded cells for the spectrochemical analysis of radioisotope samples are described. One of them is devoted to the evaporation of samples before excitation and the other one contains a suitable spectrographic excitation stand for the copper spark technique. A special device makes it possible the easy displacement of the excitation cell on wheels and rails for its accurate and reproducible position as well as its replacement by a glove box for plutonium analysis. In order to guarantee safety the room in which the spectrograph and the source are set up in separated from the active laboratory by a wall with a suitable window. (Author) 1 refs

  17. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    International Nuclear Information System (INIS)

    Padula, D.

    2000-01-01

    The scope of this analysis is to estimate the shielding wall, ceiling or equivalent door thicknesses that will be required in the Waste Handling Building to maintain the radiation doses to personnel within acceptable limits. The shielding thickness calculated is the minimum required to meet administrative limits, and not necessarily what will be recommended for the final design. The preliminary evaluations will identify the areas which have the greatest impact on mechanical and facility design concepts. The objective is to provide the design teams with the necessary information to assure an efficient and effective design

  18. 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

  19. Status of reactor shielding research in the United States

    International Nuclear Information System (INIS)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties

  20. Facility design: introduction

    International Nuclear Information System (INIS)

    Unger, W.E.

    1980-01-01

    The design of shielded chemical processing facilities for handling plutonium is discussed. The TRU facility is considered in particular; its features for minimizing the escape of process materials are listed. 20 figures

  1. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    Westbrook, J.L.; DeVore, J.R.

    1995-08-01

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  2. Program for photon shielding calculations. Examination of approximations on irradiation geometries

    International Nuclear Information System (INIS)

    Isozumi, Yasuhito; Ishizuka, Fumihiko; Miyatake, Hideo; Kato, Takahisa; Tosaki, Mitsuo

    2004-01-01

    Penetration factors and related numerical data in 'Manual of Practical Shield Calculation of Radiation Facilities (2000)', which correspond to the irradiation geometries of point isotropic source in infinite thick material (PI), point isotropic source in finite thick material (PF) and vertical incident to finite thick material (VF), have been carefully examined. The shield calculation based on the PI geometry is usually performed with effective dose penetration factors of radioisotopes given in the 'manual'. The present work cleary shows that such a calculation may lead to an overestimate more than twice larger, especially for thick shield of concrete and water. Employing the numerical data in the 'manual', we have fabricated a simple computer program for the estimation of penetration factors and effective doses of radioisotopes in the different irradiation geometries, i.e., PI, PF and VF. The program is also available to calculate the effective dose from a set of radioisotopes in the different positions, which is necessary for the γ-ray shielding of radioisotope facilities. (author)

  3. Re-evaluation of structural shielding designs of X-ray and CO-60 gamma-ray scanners at the Port of Tema, Ghana

    International Nuclear Information System (INIS)

    Ofori, K.

    2011-07-01

    This research work was conducted to re-evaluate the shielding designs of the 6 MeV x-ray and the 1.253 MeV Co-60 gamma ray scanners used for cargo-containerized scanning at the Port of Tema. These scanners utilize ionizing radiation, therefore adequate shielding must be provided to reduce the radiation exposure of persons in and around the facilities to acceptable levels. The purpose of radiation shielding is to protect workers and the general public from the harmful effects of ionizing radiation. Investigations on the facilities indicated that after commissioning, no work had been carried out to re-evaluate the shielding designs. However, workloads have increased over time neccessitating review of the installed shielding. There has been introduction of scanner units with higher radiation energy (as in the case of the x-ray scanner) posibily increasing dose rates at various location requiring review of the shielding. New structures have been dotted around the facilities without particular attention to their distances and locations with respect to the radiation source. Measurements of distances from the source axes to the points of concern for primary and leakage barrier shielding; source to container and container to the points of concern for scattered radiation shielding were taken. The primary and secondary thicknesses required for both scanners were determined based on current operational parameters and compared with the thickness constituted during the construction of the facilities. Calculated and measured dose rate beyond the shielding barriers were used to established the adequacy or otherwise of the shielding employed by the shielding designers. Values obtained fell below the 20 µSv/hr specified by NCRP 151 (2005) which showed that the primary and secondary shields of both facilities were adequate requiring no additional shielding. (author)

  4. Validation of Shielding Analysis Capability of SuperMC with SINBAD

    Directory of Open Access Journals (Sweden)

    Chen Chaobin

    2017-01-01

    Full Text Available Abstract: The shielding analysis capability of SuperMC was validated with the Shielding Integral Benchmark Archive Database (SINBAD. The SINBAD was compiled by RSICC and NEA, it includes numerous benchmark experiments performed with the D-T fusion neutron source facilities of OKTAVIAN, FNS, IPPE, etc. The results from SuperMC simulation were compared with experimental data and MCNP results. Very good agreement with deviation lower than 1% was achieved and it suggests that SuperMC is reliable in shielding calculation.

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. Design Report for ACP Hot Cell Rear Door

    Energy Technology Data Exchange (ETDEWEB)

    Ku, J. H.; Kwon, K. C.; Choung, W. M.; Cho, I. J.; Kook, D. H.; Lee, W. K.; You, G. S.; Lee, E. P.; Park, S. W

    2005-12-15

    A hot-cell facility was constructed at the IMEF building for the demonstrate ACP process. ACP hot-cell consists of process cell and maintenance cell, and each cell has rear door. Since this facility was constructed at basement floor, all process materials, equipment and radioactive materials are take in and out through the rear door. Also, this door can be an access route of workers for the maintenance works. Therefore ACP hot-cell rear doors must maintain the radiation shielding, sealing, mechanical and structural safety. This report presents design criteria, design contents of each part and driving part. It was confirmed that the rear doors sufficiently maintain the safety through the structural analysis and shielding analysis. Also, it was confirmed that the rear doors were constructed as designed by the gamma scanning test after the installation.

  10. Design Report for ACP Hot Cell Rear Door

    International Nuclear Information System (INIS)

    Ku, J. H.; Kwon, K. C.; Choung, W. M.; Cho, I. J.; Kook, D. H.; Lee, W. K.; You, G. S.; Lee, E. P.; Park, S. W.

    2005-12-01

    A hot-cell facility was constructed at the IMEF building for the demonstrate ACP process. ACP hot-cell consists of process cell and maintenance cell, and each cell has rear door. Since this facility was constructed at basement floor, all process materials, equipment and radioactive materials are take in and out through the rear door. Also, this door can be an access route of workers for the maintenance works. Therefore ACP hot-cell rear doors must maintain the radiation shielding, sealing, mechanical and structural safety. This report presents design criteria, design contents of each part and driving part. It was confirmed that the rear doors sufficiently maintain the safety through the structural analysis and shielding analysis. Also, it was confirmed that the rear doors were constructed as designed by the gamma scanning test after the installation

  11. Vertical removable filters in shielded casing for radioactive cells and process gaseous wastes

    International Nuclear Information System (INIS)

    Prinz, M.

    1983-01-01

    The installation of shielded filtration casing is necessary for highly contaminated active cells and process gaseous wastes containing active aerosols. SGN and COGEMA have developed two filtration casings (for 500 and 3000 m 3 /h flow rates) equipped with a vertically removable filter element. The filter elements fitted with high efficiency glass fiber media, are cylindrical in shape. The top flange of the filter is equipped with a gasket to ensure sealing between the filter element and its casing. The filter element is blindly installed and removed and its orientation, inside the casing, is immaterial. The shielding casing is made of a cast iron, or steel, shielding slab under which is secured the filtration casing itself. This shielding slab is settled on side shielding walls made of concrete or cast iron. The filter element, integral with a plug, is placed in the horizontal slab. The attachment of the filter element under the plug is necessary so that the plug and filter may be removed as one unit, and to keep the filter on its sealing surfaces, according to sealing and seismic resistance requirements. Filter removal is performed with the help of an intervention cask, centered over a removable trap door provided on the shielding slab of the casing. First, the plug and filter element assembly is raised into the cask. Then, the filtering element may be separated from the plug which is decontaminated and salvaged. The whole plug and filter assembly may also be sent to the conditioning waste storage. The installation of a clean filter element in the casing, is also performed with the help of the intervention cask, proceeding as above, but in reverse order. The same intervention cask may also be used to remove the upstream and downstream dampers from the top of the casing

  12. Assessment of the structural shielding integrity of some selected computed tomography facilities in the Greater Accra Region of Ghana

    International Nuclear Information System (INIS)

    Nkansah, A.

    2010-01-01

    The structural shielding integrity was assessed for four of the CT facilities at Trust Hospital, Korle-Bu Teaching Hospital, the 37 Military Hospital and Medical Imaging Ghana Ltd. in the Greater Accra Region of Ghana. From the shielding calculations, the concrete wall thickness computed are 120, 145, 140 and 155mm, for Medical Imaging Ghana Ltd. 37 Military, Trust Hospital and Korle-Bu Teaching Hospital respectively using Default DLP values. The wall thickness using Derived DLP values are 110, 110, 120 and 168mm for Medical Imaging Ghana Ltd, 37 Military Hospital, Trust Hospital and Korle-Bu Teaching Hospital respectively. These values are within the accepted standard concrete thickness of 102- 152mm prescribed by the National Council of Radiological Protection and measurement. The ultrasonic pulse testing indicated that all the sandcrete walls are of good quality and free of voids since pulse velocities estimated were approximately equal to 3.45km/s. an average dose rate measurement for supervised areas is 3.4 μSv/wk and controlled areas is 18.0 μSv/wk. These dose rates were below the acceptable levels of 100 μSv per week for the occupationally exposed and 20 μSv per week for members of the public provided by the ICRU. The results mean that the structural shielding thickness are adequate to protect members of the public and occupationally exposed workers (au).

  13. Bulk shielding facility semi-annual report, January--June 1990

    International Nuclear Information System (INIS)

    Laughlin, D.L.; Coleman, G.H.

    1990-11-01

    The Bulk Shielding Reactor (BSR) remained shut down during January, February, March, April, May, and June. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. Maintenance and changes are described. The Pool Critical Assembly (PCA) remains shut down, but surveillance tests are described

  14. Shielding consideration for a deuteron activated liquid lithium system

    International Nuclear Information System (INIS)

    Huang, S.T.; Shapiro, A.M.; Lee, J.B.; Miller, W.C.

    1979-09-01

    A parametric study was conducted to evaluate the potential shielding implication due to the 7 Be plateout on the lithium piping in the FMIT facility. Various parameters such as plateout percentage, hot flush efficiency and 7 Be trapping efficiency were varied to assess the overall shielding requirement relationship. The 7 Be plateout was found to place severe limitations on the hands-on maintenance access. Hot flush and 7 Be traps are effective ways of minimizing the 7 Be plateout. To be effective in reducing local shielding requirements, the combined 7 Be trapping and hot flush efficiency shall be greater than 95%

  15. SU-E-T-270: Optimized Shielding Calculations for Medical Linear Accelerators (LINACs).

    Science.gov (United States)

    Muhammad, W; Lee, S; Hussain, A

    2012-06-01

    The purpose of radiation shielding is to reduce the effective equivalent dose from a medical linear accelerator (LINAC) to a point outside the room to a level determined by individual state/international regulations. The study was performed to design LINAC's room for newly planned radiotherapy centers. Optimized shielding calculations were performed for LINACs having maximum photon energy of 20 MV based on NCRP 151. The maximum permissible dose limits were kept 0.04 mSv/week and 0.002 mSv/week for controlled and uncontrolled areas respectively by following ALARA principle. The planned LINAC's room was compared to the already constructed (non-optimized) LINAC's room to evaluate the shielding costs and the other facilities those are directly related to the room design. In the evaluation process it was noted that the non-optimized room size (i.e., 610 × 610 cm 2 or 20 feet × 20 feet) is not suitable for total body irradiation (TBI) although the machine installed inside was having not only the facility of TBI but the license was acquired. By keeping this point in view, the optimized INAC's room size was kept 762 × 762 cm 2. Although, the area of the optimized rooms was greater than the non-planned room (i.e., 762 × 762 cm 2 instead of 610 × 610 cm 2), the shielding cost for the optimized LINAC's rooms was reduced by 15%. When optimized shielding calculations were re-performed for non-optimized shielding room (i.e., keeping room size, occupancy factors, workload etc. same), it was found that the shielding cost may be lower to 41 %. In conclusion, non- optimized LINAC's room can not only put extra financial burden on the hospital but also can cause of some serious issues related to providing health care facilities for patients. © 2012 American Association of Physicists in Medicine.

  16. Evaluation of the room shielding thickness of Hi-Art tomotherapy system

    International Nuclear Information System (INIS)

    Liu Haikuan; Wu Jinhai; Gu Naigu; Gao Yiming; Wang Li; Huang Weiqin; Wang Fengxian

    2010-01-01

    In this paper, we calculate and evaluate the room shielding thickness of a Hi-Art tomotherapy system, which is a new type of radiotherapy facility. Due to the self-shielding of the accelerator,only scattered beam and beam leakage were considered in calculating the room shielding thickness. The radiation field of the tomotherapy system was used as the basic data to calculate the shielding thickness of every 15 degree solid angle. The maximum shielding thickness required of each shielding wall was at the position with the angle of 15 degree, and the calculated shielding thickness were 1023, 975, 917, 1460, 1147 and 1189 mm for the east wall,south wall,west wall, north wall, the roof and the floor,respectively. According to the calculation results, all shielding walls, ceiling and floor could meet the requirement of the radiation protection, but the north wall thickness of 1200 mm was a little thinner. (authors)

  17. Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

    International Nuclear Information System (INIS)

    Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E; Wagner, John C; McMahan, Kimberly L; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Trama, Jean-Christophe; Masse, Veronique; Gagnier, Emmanuel; Naury, Sylvie; Blanc-Tranchant, Patrick; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

    2015-01-01

    In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available

  18. 2015 In-Situ Gamma-Ray Assay of the West Cell Line in the 235-F Plutonium Fuel Form Facility

    Energy Technology Data Exchange (ETDEWEB)

    Brand, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Aucott, T. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DiPrete, D. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-02-01

    In November and December 2015, scientists from SRNL took a series of in-situ gamma-ray measurements through the windows in front of Cells 6-9 on the west line of the PuFF facility using a shielded, 120% high-purity germanium detector. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Where possible, the distribution of the Pu-238 in the cells was determined using the Germanium Gamma-ray Imager (GeGI). This distribution was then fed into the MCNP model to quantify the Pu-238 in each cell. Data analysis was performed using three gamma rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells.

  19. Cost and schedule estimate to construct the tunnel and shaft remedial shielding concept, Los Alamos Meson Physics Facility, Los Alamos National Laboratory, Los Alamos, New Mexico. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-11-30

    The report provides an estimate of the cost and associated schedule to construct the tunnel and shaft remedial shielding concept. The cost and schedule estimate is based on a preliminary concept intended to address the potential radiation effects on Line D and Line Facilities in event of a beam spill. The construction approach utilizes careful tunneling methods based on available excavation and ground support technology. The tunneling rates and overall productivity on which the cost and project schedule are estimated are based on conservative assumptions with appropriate contingencies to address the uncertainty associated with geological conditions. The report is intended to provide supplemental information which will assist in assessing the feasibility of the tunnel and shaft concept and justification for future development of this particular aspect of remedial shielding for Line D and Line D Facilities.

  20. 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

  1. The shielded chemistry suite at BNL and its use in the study of the irradiated uranium-water reaction

    International Nuclear Information System (INIS)

    Bannister, G.H.; Tyfield, S.P.

    1977-11-01

    A versatile shielded facility for chemical studies is now in use at Berkeley Nuclear Laboratories. The suite was fully commissioned in 1976 and first utilised to study the aqueous corrosion of irradiated reactor grade uranium. The suite consists of three adjacent interconnected cells; a balance cell, a preparation cell and an experimental cell. The development, design and layout of the facilities are described, as are the details of the equipment installed to undertake a uranium corrosion study. The preparation cell is used for procedures that could results in high levels of contamination, such as cutting of uranium specimens. The uranium samples are corroded in sealed steel capsules installed in the experimental cell. The capsules are heated either isothermally or through a transient and the reaction monitored with a pressure transducer which records the rate of hydrogen evolution from the uranium-water reaction. The experimental results will be reported elsewhere. (author)

  2. Castor and Pollux - shielded cells for studying fuel treatment processes

    International Nuclear Information System (INIS)

    Faudot, G.; Bathellier, A.

    1969-01-01

    CASTOR and POLLUX, two alpha, beta, gamma cells are described in the present paper. They are located in the CEN at Fontenay-aux-Roses (France). They are designed for improvement studies of the various aqueous separation processes used in irradiated fuels reprocessing plants. Located in the same air-tight steel encasement, they arc inter-connected by a pneumatic transfer. These two cells have a similar in-line conception and they include: a gamma shielding in lead of 10 cm of thickness; an inner air-tight box, made with stainless steel and plexiglas, is maintained in lowering in comparison to room pressure. Eleven Hobson model seven master-slave manipulators allow inner manipulations. Then the inner equipment is described briefly. (author) [fr

  3. Validation of PHITS Spallation Models from the Perspective of the Shielding Design of Transmutation Experimental Facility

    Science.gov (United States)

    Iwamoto, Hiroki; Meigo, Shin-ichiro

    2017-09-01

    The impact of different spallation models implemented in the particle transport code PHITS on the shielding design of Transmutation Experimental Facility is investigated. For 400-MeV proton incident on a lead-bismuth eutectic target, an effective dose rate at the end of a thick radiation shield (3-m-thick iron and 3-m-thick concrete) calculated by the Liège intranuclear cascade (INC) model version 4.6 (INCL4.6) coupled with the GEMcode (INCL4.6/GEM) yields about twice as high as the Bertini INC model (Bertini/GEM). A comparison with experimental data for 500-MeV proton incident on a thick lead target suggest that the prediction accuracy of INCL4.6/GEM would be better than that of Bertini/GEM. In contrast, it is found that the dose rates in beam ducts in front of targets calculated by the INCL4.6/GEMare lower than those by the Bertini/GEM. Since both models underestimate the experimental results for neutron-production doubledifferential cross sections at 180° for 140-MeV proton incident on carbon, iron, and gold targets, it is concluded that it is necessary to allow a margin for uncertainty caused by the spallation models, which is a factor of two, in estimating the dose rate induced by neutron streaming through a beam duct.

  4. ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

  5. Closure Report for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada National Security Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Restoration

    2011-09-29

    This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 [as amended March 2010]). CAU 116 consists of the following two Corrective Action Sites (CASs), located in Area 25 of the Nevada National Security Site: (1) CAS 25-23-20, Nuclear Furnace Piping and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 consisted of Building 3210 and the attached concrete shield wall. CAS 25-23-20 consisted of the nuclear furnace piping and tanks. Closure activities began in January 2007 and were completed in August 2011. Activities were conducted according to Revision 1 of the Streamlined Approach for Environmental Restoration Plan for CAU 116 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2008). This CR provides documentation supporting the completed corrective actions and provides data confirming that closure objectives for CAU 116 were met. Site characterization data and process knowledge indicated that surface areas were radiologically contaminated above release limits and that regulated and/or hazardous wastes were present in the facility.

  6. Shielding features of quarry stone

    International Nuclear Information System (INIS)

    Hernandez V, C.; Contreras S, H.; Hernandez A, L.; Baltazar R, A.; Escareno J, E.; Mares E, C. A.; Vega C, H. R.

    2010-10-01

    Quarry stone lineal attenuation coefficient for gamma-rays has been obtained. In Zacatecas, quarry stone is widely utilized as a decorative item in buildings, however its shielding features against gamma-rays unknown. The aim of this work is to determine the shielding properties of quarry stone against γ-rays using Monte Carlo calculations where a detailed model of a good geometry experimental setup was carried out. In the calculations 10 pieces 10 X 10 cm 2 of different thickness were utilized to evaluate the photons transmission as the quarry stone thickness is increased. It was noticed that transmitted photons decay away as the shield thickness is increased, these results were fitted to an exponential function were the linear attenuation coefficient was estimated. Also, using XCOM code the linear attenuation coefficient from several keV up to 100 MeV was estimated. From the comparison between Monte Carlo results and XCOM calculations a good agreement was found. For 0.662 MeV γ-rays the attenuation coefficient of quarry stone, whose density is 2.413 g-cm -3 , is 0.1798 cm -1 , this mean a X 1/2 = 3.9 cm, X 1/4 = 7.7 cm, X 1/10 = 12.8 cm, and X 1/100 = 25.6 cm. Having the information of quarry stone performance as shielding give the chance to use this material to shield X and γ-ray facilities. (Author)

  7. Shielding required for radiation produced by 15 GeV stored electrons

    International Nuclear Information System (INIS)

    Jenkins, T.M.; McCaslin, J.B.; Thomas, R.H.

    1974-01-01

    The first phase of PEP will consist of a 15 GeV electron and positron storage ring. This note examines the shielding required by such a facility. Shielding of neutrons and muons produced by 200 GeV protons in the second phase of PEP has been discussed in previous notes. 9 refs., 9 figs., 5 tabs

  8. 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

  9. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Michael R. Kruzic

    2007-01-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility was used in the early to mid-1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles in the immediate area. Identified as Corrective Action Unit 115, the TCA facility was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the ''Federal Facility Agreement and Consent Order''. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously provided technical decisions are made by an experienced decision maker within the site conceptual site model, identified in the Data Quality Objective process. Facility closure involved a seven-step decommissioning strategy. Key lessons learned from the project included: (1) Targeted preliminary investigation activities provided a more solid technical approach, reduced surprises and scope creep, and made the working environment safer for the D and D worker. (2) Early identification of risks and uncertainties provided opportunities for risk management and mitigation planning to address challenges and unanticipated conditions. (3) Team reviews provided an excellent mechanism to consider all aspects of the task, integrated safety into activity performance, increase team unity and ''buy-in'' and promoted innovative and time saving ideas. (4) Development of CED protocols ensured safety and control. (5) The same proven D and D strategy is now being employed on the larger ''sister'' facility, Test Cell C

  10. 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

  11. Shielding of the abdominal region during X-irradiation: Effect on haemopoietic stem cells

    International Nuclear Information System (INIS)

    Vavrova, J.; Petyrek, P.

    1984-01-01

    The shielding of the abdominal region during X-irradiation is important for two reasons: 1) it prevents the development of the gastrointestinal syndrome following higher radiation doses; 2) it prevents the development of the lethal form of the bone marrow syndrome by shielding a portion of the spinal column bone marrow and the spleen. The dose reduction factor in partial irradiated mice compared to whole-body irradiated mice was 2.6. 15 mins after irradiation with the dose of 10 Gy, the number of spleen colony-forming cells (CFU-S) decreased in the femoral bone marrow of both the group receiving whole-body irradiation and the group having the abdominal region shielded. This decrease persisted in both groups for five hours after irradiation. 48 hours after whole-body irradiation, the number of CFU-S in the femoral bone marrow of the unshielded animals decreased almost to zero values. In shielded group, a statistically significant increase in the number of CFU-S in the femur was observed at 48 hours. At 120 hours, the number of CFU-S in so irradiated mice was at the level of the unirradiated group. In the spleen a great decrease in the number of CFU-S was caused by whole-body irradiation of mice as early as 15 mins after irradiation and the low-levels were recorded until the death of animals. In the shielded group, a statistically signifjcant decrease in the number of CFU-S was observed 15 mins after irradiation, which may have been due to the abscopal effect of irradiation. After a transient recovery 2 hours after irradiation, a significant decrease in the number of CFU-S in the spleens was observed in the subsequent interval from five to 120 hours after irradiation. (author)

  12. FLUKA shielding calculations for the FAIR project

    International Nuclear Information System (INIS)

    Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey

    2015-01-01

    FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.

  13. Technical study of a thermally dense long term interim storage facility

    International Nuclear Information System (INIS)

    Le Duigou, A.; Badie, M.; Duret, B.; Bricard, A.

    2001-01-01

    The COFRE concept is aimed at the surface and thermal densification of the interim storage facility for irradiated fuels. The facility provides the biological shielding. A conditioning cell is used to load and retrieve the fuel assemblies. The facility container is the second containment barrier. The high power levels are managed by an auxiliary cooling system whose original feature is the passive use of a water evaporation-condensation cycle in a sealed circuit. The removable evaporator abuts the container. The air cooled condenser is placed outside the facility. Contact resistance and heat pipe mode were successfully modelled and are undergoing experimental validation on the THERESE and REBECA loops. (author)

  14. Cost benefit analysis of the radiological shielding of medical cyclotrons using a genetic algorithm

    International Nuclear Information System (INIS)

    Mukherjee, Bhaskar

    2001-01-01

    Adequate radiation shielding is vital to the safe operation of modern commercial medical cyclotrons producing large yields of short-lived radioisotopes. The radiological shielding constitutes a significant capital investment for any new cyclotron-based radioisotope production facility; hence, the shielding design requires an accurate cost-benefit analysis often based on a complex multi-variant optimization technique. This paper demonstrates the application of a Genetic Algorithm (GA) for the optimum design of the high yield target cave of a Medical Cyclotron radioisotope production facility based in Sydney, Australia. The GA is a novel optimization technique that mimics the Darwinian Evolution paradigm and is ideally suited to search for global optima in a large multi-dimensional solution space

  15. Evaluation of the Shielding Performance for the Hot-cell built in 100-MeV Isotope Beam-line of KOMAC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Min; Park, Sung Kyun; Min, Yi Sub; Cho, Yong Sub [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study describes the structure of the hot-cell constructed in KOMAC for radioisotope production and evaluates the shielding performance for the hot-cell via the radiation shielding ability test. Korea multi-purpose accelerator complex (KOMAC) is currently operating 20-MeV and 100-MeV beam-line one by on. Additional 100-MeV beam-line and target room (TR101) are planned for the purpose of the radioisotope production in this year. The initial goal of the radioisotope production is to produce the radioactive isotopes, Sr-82 or Cu-67, used widely for the diagnosis and treatment of the cancer. In order to produce these radioisotopes mentioned, the proton beam with the energy between 70-MeV and 100- MeV at a beam current of 300 μA is irradiated into a solid target made of ZnO or RbCl. After the irradiation of the proton beam during approximately 100 hours, the radioisotope Sr-82 with the radioactivity amount of about 3.8 Ci or the Cu-67 with the amount of about 2.7 Ci will be produced. Radioisotopes produced though this process should be conveyed from the TR101 target room to the PR101 processing room and then in order to be delivered into the place for the next process step, a hot-cell is necessary. Result of the shielding performance evaluation of the hot-cell for producing radioisotopes shows the necessity of the shield reinforcement using lead material at side of the lead glass window.

  16. Accelerator shielding experts meet at CERN

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    Fifteen years after its first CERN edition, the Shielding Aspects of Accelerator, Targets and Irradiation Facility (SATIF) conference was held again here from 2-4 June. Now at its 10th edition, SATIF10 brought together experts from all over the world to discuss issues related to the shielding techniques. They set out the scene for an improved collaboration and discussed novel shielding solutions.   This was the most attended meeting of the series with more than 65 participants from 34 institutions and 14 countries. “We welcomed experts from many different laboratories around the world. We come from different contexts but we face similar problems. In this year’s session, among other things, we discussed ways for improving the effectiveness of calculations versus real data, as well as experimental solutions to investigate the damage that radiation produces on various materials and the electronics”, says Marco Silari, Chair of the conference and member of the DGS/RP gro...

  17. Shielding measurements for a 230 MeV proton beam

    International Nuclear Information System (INIS)

    Siebers, J.V.

    1990-01-01

    Energetic secondary neutrons produced as protons interact with accelerator components and patients dominate the radiation shielding environment for proton radiotherapy facilities. Due to the scarcity of data describing neutron production, attenuation, absorbed dose, and dose equivalent values, these parameters were measured for 230 MeV proton bombardment of stopping length Al, Fe, and Pb targets at emission angles of 0 degree, 22 degree, 45 degree, and 90 degree in a thick concrete shield. Low pressure tissue-equivalent proportional counters with volumes ranging from 1 cm 3 to 1000 cm 3 were used to obtain microdosimetric spectra from which absorbed dose and radiation quality are deduced. Does equivalent values and attenuation lengths determined at depth in the shield were found to vary sharply with angle, but were found to be independent of target material. Neutron dose and radiation length values are compared with Monte Carlo neutron transport calculations performed using the Los Alamos High Energy Transport Code (LAHET). Calculations used 230 MeV protons incident upon an Fe target in a shielding geometry similar to that used in the experiment. LAHET calculations overestimated measured attenuation values at 0 degree, 22 degree, and 45 degree, yet correctly predicted the attenuation length at 90 degree. Comparison of the mean radiation quality estimated with the Monte Carlo calculations with measurements suggest that neutron quality factors should be increased by a factor of 1.4. These results are useful for the shielding design of new facilities as well as for testing neutron production and transport calculations

  18. 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.

  19. Shielding research in France

    Energy Technology Data Exchange (ETDEWEB)

    Lafore, P

    1964-10-01

    Shielding research as an independent subject in France dates from 1956. The importance of these studies has been reflected in the contribution which they have made to power reactor design and in the resultant savings in expenditure for civil engineering and machinery for the removal of mobile shields. The Reactor Shielding Research Division numbers approximately 60 persons and uses several experimental facilities. These include: NAIADE I, installed near the ZOE reactor and operating with a natural uranium slab 2 cm thick (an effective diameter of 60 cm is the one most commonly used); the TRITON pool-type reactor, mainly used in shielding studies, includes an active-water loop, by means of which the secondary shields required for light-water reactors can be studied; core, NEREIDE, which is situated near a 2 m x 2 m aluminium window enables a large neutron source to be placed in a compartment without water in which large-scale mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavities, and of reactor shielding of greater thickness than that in NAIADE I; SAMES 600 keV accelerator is used for monoenergetic neutron studies. Instrumentation studies are an important part of the work, mainly in the measurement of fast neutrons and their spectra by activation detectors. Of late, attention has been directed towards the use of (n, n') (rhodium) reactions and of heavy detectors for low-flux measurements. The simultaneous use of a large number of detectors poses automation problems. With our installation we can count 16 detectors simultaneously. Neutron spectrum studies are conducted with nuclear emulsions and a lithium-6 semiconductor spectrometer. As to the materials used, the research carried out in France involves chiefly graphite, iron and concrete at various temperatures up to 800 deg C. Different compounds, borated and non-borated and of densities up to between 1 and 9 are under consideration. Problems connected with applications are

  20. Advanced facilities for radiochemistry at Harwell

    International Nuclear Information System (INIS)

    1985-01-01

    The leaflets in this folder describe the latest addition to Harwell's active handling capability. This is a high level alpha, beta, gamma facility designed specifically for undertaking chemical research and development work. It is based on using high integrity containment boxes which are housed in concrete shielded enclosures. The active boxes can be removed and transferred remotely to a support area where they, and any associated equipment, can be decontaminated and serviced whilst a new fully commissioned box can be readily brought into service. The facility fulfills the principle of ALARA and is sufficiently flexible to accommodate a wide range of active handling requirements. It is supported by a suite of medium active handling cells, radiochemical laboratories and, as necessary, facilities of other scientific and engineering disciplines. The leaflets are: report on conceptual aspects; Techsheet 'Remote handling facility - Salient information'; Techsheet 'Project capabilities'; and 4 sheets of diagrams showing details of the facility. (U.K.)

  1. 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.

  2. Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Nojiri, Ichiro; Odajima, Akira; Sasaki, Toshihisa; Kurosawa, Naohiro

    1998-01-01

    In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)

  3. Simplified shielding calculation system for high-intensity proton accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Masumura, Tomomi; Nakashima, Hiroshi; Nakane, Yoshihiro; Sasamoto, Nobuo [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-06-01

    A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)

  4. A shielding chamber for the Rossendorf whole body counter

    International Nuclear Information System (INIS)

    Beutmann, A.; Ebert, S.; Kaden, M.; Loehnert, D.; Doerfel, H.R.; Schreiber, W.; Helbig, S.

    2016-01-01

    In connection with the relocation of the incorporation measurement point operated by the VKTA, a new shielding chamber was designed. The development of the new shielding chamber will be shown based on the design study by IDEA System and the inquiries for material availability, procurement of material and assembly technology up to fabrication, assembly and completion of the chamber. The accompanying background measurements through In-situ gamma spectrometry and first experiences with incorporation measurements at the new In-Vivo measurement facility are shown.

  5. 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

  6. Fabrication and electromagnetic interference shielding performance of open-cell foam of a Cu–Ni alloy integrated with CNTs

    Energy Technology Data Exchange (ETDEWEB)

    Ji, Keju; Zhao, Huihui; Zhang, Jun; Chen, Jia; Dai, Zhendong, E-mail: zddai@nuaa.edu.cn

    2014-08-30

    Highlights: • Cu–Ni alloy open-cell foam integrated with CNTs was used for EMI shielding. • The composite was prepared by electroless, electro-, and electrophoretic deposition. • The main shielding mechanism was multiple reflections and absorptions of microwaves. • The composite had a porous structure, large surface area, and inherent permeability. - Abstract: A lightweight multi-layered electromagnetic interference (EMI) shielding material made of open-cell foam of a Cu–Ni alloy integrated with carbon nanotubes (CNTs) was prepared by electroless copper plating, then nickel electroplating, and finally electrophoretic deposition of CNTs. The foamed Cu–Ni–CNT composite comprises, from inside to outside, Cu, Ni, and CNT layers. Scanning electron microscopy, energy dispersive spectroscopy, and EMI tests were employed to characterize the morphology, composition, and EMI performance of the composite, respectively. The results indicated that the shielding effectiveness (SE) of the composite increased with increasing pore density (indicated as pores per inch (PPI)) and increasing thickness. A specimen with a PPI of 110 and a 1.5-mm thickness had a maximum SE of up to 54.6 dB, and a SE as high as 47.5 dB on average in the 8–12 GHz range. Integrating the inherent superiority of Cu, Ni, and CNTs, the porous structure of the composite can attenuate the incident electromagnetic microwaves by reflecting, scattering, and absorbing them between the metallic skeleton and the CNT layer. The multiple reflections and absorptions make it difficult for the microwaves to escape from the composite before being absorbed, thereby making the composite a potential shielding material.

  7. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  8. Concrete mix design for X-and gamma shielding

    International Nuclear Information System (INIS)

    Mohamad Pauzi Ismail; Noor Azreen Masenwat; Suhairy Sani; Abdul Bakhri Muhammad; Mohd Kamal Shah Shamsuddin; Rahmad Abd Rashid

    2012-01-01

    The design of X-ray or gamma ray radiographic exposure room requires some calculations on shielding to provide safe operation of the facility and minimum exposure to radiation workers. Careful design can lead to economical installations with minimal barriers. The design depends on such factors as: maximum energy, maximum intensity, permitted full-body dosage, workload, use factor, occupancy factor, maximum dose output and shielding materials. Choice of material for a barrier depends on convenience and cost. The radiographic exposure room is usually made of normal concrete with density of about 2.3 - 2.4 g/ cc. Normal concrete is often used for construction of exposure room because of cheap and ease of construction. This paper explained and discussed the optimum mix design for normal concrete used for X-and gamma shielding. (author)

  9. Concept of spatial channel theory applied to reactor shielding analysis

    International Nuclear Information System (INIS)

    Williams, M.L.; Engle, W.W. Jr.

    1977-01-01

    The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield

  10. 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

  11. Application of the characteristics method combined with advanced self-shielding models to an ACR-type cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Hebert, A.

    2005-01-01

    In this paper, we present the usage of the method of characteristics (MOC) with advanced self-shielding models for a fundamental lattice calculation on an ACR-type cell i.e. a cluster geometry with light water coolant and heavy water moderator. Comparison with the collision probability method (CP) show the consistency of the method of characteristics as implemented both in flux and self-shielding calculations. Acceleration techniques are tested in the different calculations and prove to be efficient. Comparisons with the Monte-Carlo code Tripoli4 show the advantage of a subgroup approach for self-shielding calculations : the difference in k eff is less than one standard deviation of the Tripoli4 calculation and in terms of total absorption rates, in the resolved resonances group, the maximum relative error is of the order of 3% localised in the most outer region of the central pin. (author)

  12. Preliminary shielding analysis of VHTR reactors

    International Nuclear Information System (INIS)

    Flaspoehler, Timothy M.; Petrovic, Bojan

    2011-01-01

    Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

  13. Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

    International Nuclear Information System (INIS)

    Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

    2003-01-01

    The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

  14. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  15. Benchmarking shielding simulations for an accelerator-driven spallation neutron source

    Directory of Open Access Journals (Sweden)

    Nataliia Cherkashyna

    2015-08-01

    Full Text Available The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS, currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ, at the Paul Scherrer Institute (PSI, Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters instruments at ESS.

  16. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility; Monte-Carlo-basierter Nachweis der ausreichend dimensionierten Abschirmung einer {sup 60}Co-Pruefanlage

    Energy Technology Data Exchange (ETDEWEB)

    Wind, Michael; Beck, Peter; Latocha, Marcin [Seibersdorf Labor GmbH, Radiation Protection Dosimetry, Seibersdorf (Austria)

    2015-07-01

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  17. Grounding and shielding circuits and interference

    CERN Document Server

    Morrison, Ralph

    2016-01-01

    Applies basic field behavior in circuit design and demonstrates how it relates to grounding and shielding requirements and techniques in circuit design This book connects the fundamentals of electromagnetic theory to the problems of interference in all types of electronic design. The text covers power distribution in facilities, mixing of analog and digital circuitry, circuit board layout at high clock rates, and meeting radiation and susceptibility standards. The author examines the grounding and shielding requirements and techniques in circuit design and applies basic physics to circuit behavior. The sixth edition of this book has been updated with new material added throughout the chapters where appropriate. The presentation of the book has also been rearranged in order to reflect the current trends in the field.

  18. PROMPT DOSE ANALYSIS FOR THE NATIONAL IGNITION FACILITY

    International Nuclear Information System (INIS)

    Khater, H.; Dauffy, L.; Sitaraman, S.; Brereton, S.

    2008-01-01

    Detailed 3-D modeling of the NIF facility is developed to accurately understand the prompt radiation environment within NIF. Prompt dose values are calculated for different phases of NIF operation. Results of the analysis were used to determine the final thicknesses of the Target Bay (TB) and secondary doors as well as the required shield thicknesses for all unused penetrations. Integrated dose values at different locations within the facility are needed to formulate the personnel access requirements within different parts of the facility. The conclusions of this presentation are: (1) The current NIF facility model includes all important features of the Target Chamber, shielding system, and building configuration; (2) All shielding requirements for Phase I operation are met; (3) Negligible dose values (a fraction of mrem) are expected in normally occupied areas during Phase I; (4) In preparation for the Ignition Campaign and Phase IV of operation, all primary and secondary shield doors will be installed; (5) Unused utility penetrations in the Target Bay and Switchyard walls (∼50%) will be shielded by 1 foot thick concrete to reduce prompt dose inside and outside the NIF facility; (6) During Phase IV, a 20 MJ shot will produce acceptable dose levels in the occupied areas as well as at the nearest site boundary; (7) A comprehensive radiation monitoring plan will be put in place to monitor dose values at large number of locations; and (8) Results of the dose monitoring will be used to modify personnel access requirements if needed

  19. Shield calculations, optimization vs. paradigm; Calculos de blindajes, optimizacion vs. paradigma

    Energy Technology Data Exchange (ETDEWEB)

    Cornejo D, N.; Hernandez S, A.; Martinez G, A. [Centro de Proteccion e Higiene de las Radiaciones, Calle 20 No. 4113 e/41 y 47 Playa C.P. 11300 LaHabana (Cuba)]. e-mail: nestor@cphr.edu.cu

    2006-07-01

    Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of {mu}Sv.h{sup -1}, independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)

  20. 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.)

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. Shielding design for the target room of the proton accelerator research center

    International Nuclear Information System (INIS)

    Min, Y. S.; Lee, C. W.; Mun, K. J.; Nam, J.; Kim, J. Y.

    2010-01-01

    The Proton Engineering Frontier Project (PEFP) has been developing a 100-MeV proton linear accelerator. Also, PEFP has been designing the Proton Accelerator Research Center (PARC). In the Accelerator Tunnel and Beam Experiment Hall in PARC, 10 target rooms for the 20- and 100-MeV beamline facilities exist in the Beam Experiment Hall. For the 100-MeV target rooms during 100-MeV proton beam extraction, a number of high energy neutrons, ranging up to 100-MeV, are produced. Because of the high beam current and space limitations of each target room, the shielding design of each target room should be considered seriously. For the shielding design of the 100-MeV target rooms of the PEFP, a permanent and removable local shield structure was adopted. To optimize shielding performance, we evaluated four different shield materials (concrete, HDPE, lead, iron). From the shielding calculation results, we confirmed that the proposed shielding design made it possible to keep the dose rate below the 'as low as reasonably achievable (ALARA)' objective.

  6. Shielding considerations for an electron linear accelerator complex for high energy physics and photonics research

    International Nuclear Information System (INIS)

    Holmes, J.A.; Huntzinger, C.J.

    1987-01-01

    Radiation shielding considerations for a major high-energy physics and photonics research complex which comprise a 50 MeV electron linear accelerator injector, a 1.0 GeV electron linear accelerator and a 1.3 GeV storage ring are discussed. The facilities will be unique because of the close proximity of personnel to the accelerator beam lines, the need to adapt existing facilities and shielding materials and the application of strict ALARA dose guidelines while providing maximum access and flexibility during a phased construction program

  7. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  8. Shielding in ungated field emitter arrays

    Energy Technology Data Exchange (ETDEWEB)

    Harris, J. R. [U.S. Navy Reserve, Navy Operational Support Center New Orleans, New Orleans, Louisiana 70143 (United States); Jensen, K. L. [Code 6854, Naval Research Laboratory, Washington, D.C. 20375 (United States); Shiffler, D. A. [Directed Energy Directorate, Air Force Research Laboratory, Albuquerque, New Mexico 87117 (United States); Petillo, J. J. [Leidos, Billerica, Massachusetts 01821 (United States)

    2015-05-18

    Cathodes consisting of arrays of high aspect ratio field emitters are of great interest as sources of electron beams for vacuum electronic devices. The desire for high currents and current densities drives the cathode designer towards a denser array, but for ungated emitters, denser arrays also lead to increased shielding, in which the field enhancement factor β of each emitter is reduced due to the presence of the other emitters in the array. To facilitate the study of these arrays, we have developed a method for modeling high aspect ratio emitters using tapered dipole line charges. This method can be used to investigate proximity effects from similar emitters an arbitrary distance away and is much less computationally demanding than competing simulation approaches. Here, we introduce this method and use it to study shielding as a function of array geometry. Emitters with aspect ratios of 10{sup 2}–10{sup 4} are modeled, and the shielding-induced reduction in β is considered as a function of tip-to-tip spacing for emitter pairs and for large arrays with triangular and square unit cells. Shielding is found to be negligible when the emitter spacing is greater than the emitter height for the two-emitter array, or about 2.5 times the emitter height in the large arrays, in agreement with previously published results. Because the onset of shielding occurs at virtually the same emitter spacing in the square and triangular arrays, the triangular array is preferred for its higher emitter density at a given emitter spacing. The primary contribution to shielding in large arrays is found to come from emitters within a distance of three times the unit cell spacing for both square and triangular arrays.

  9. Neutron shielding properties of a new high-density concrete

    International Nuclear Information System (INIS)

    Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

    2008-01-01

    The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

  10. Shielding plugs

    International Nuclear Information System (INIS)

    Makishima, Kenji.

    1986-01-01

    Purpose: In shielding plugs of an LMFBR type reactor, to restrain natural convection of heat in an annular space between a thermal shield layer and a shield shell, to prevent the lowering of heat-insulation performance, and to alleviate a thermal stress in a reactor container and the shield shell. Constitution: A ring-like leaf spring split in the direction of height is disposed in an annular space between a thermal shield layer and a shield shell. In consequence, the space is partitioned in the direction of height and, therefore, if axial temperature conditions and space width are the same and the space is low, the natural convection is hard to occur. Thus the rise of upper surface temperature of the shielding plugs can prevent the lowering of the heat insulation performance which will result in the increment of shielding plug cooling capacity, thereby improving reliability. In the meantime, since there is mounted an earthquake-resisting support, the thermal shield layer will move for a slight gap in case of an earthquake, being supported by the earthquake-resisting support, and the movement of the thermal shield layer is restricted, thereby maintaining integrity without increasing the stroke of the ring-like spring. (Kawakami, Y.)

  11. Technical Requirements for Fabrication and Installation of Removable Shield for CNRF in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Shin, Jin Won

    2008-04-15

    This report details the technical requirements for the fabrication and installation of the removable shield for the Cold Neutron Research Facility (CNRF) in HANARO reactor hall. The removable shield is classified as non-nuclear safety (NNS), seismic category II, and quality class T. The main function of the removable shield is to do the biological shielding of neutrons and gamma from the CN port and the guides. The removable shield consists of block type walls and roofs that can be necessarily assembled, disassembled and moveable. These will be installed between the reactor pool wall and the CNS guide bunker in. This report describes technical requirements for the removable shield such as quality assurance, seismic analysis requirements, configuration, concrete compositions, fabrication and installation requirements, test and inspection, shipping, delivery, etc. Appendix is the technical specification of structural design and analysis. Attachments are composed of the technical specification for the fabrication of the removable shield, shielding design drawings and procurement quality requirements. These technical requirements will be provided to a contract for the manufacturing and installation.

  12. Self-shielding models of MICROX-2 code: Review and updates

    International Nuclear Information System (INIS)

    Hou, J.; Choi, H.; Ivanov, K.N.

    2014-01-01

    Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study

  13. Radiological shielding of cobalt-60 teletherapy facility at Komfo Anokye Teaching Hospital, Kumasi, Ghana

    International Nuclear Information System (INIS)

    Addison, E.K.T.; Kitcher, H.W.; Kyere, A.W.K.; Nani, E.K.

    2003-01-01

    The radiological shielding of the radiotherapy unit, oncology directorate, Komfo Anokye teaching hospital, has been assessed based on the workload and occupancy factors, in addition to the technical and performance specification of the cirus cobalt-60 equipment. Aspects relating to the design of treatment room were described. Calculations were based on NCRP report 49 and measurements were carried out on the structural shielding design of the cobalt-60 unit by the standard method prescribed in the NCRP report 49. This study was carried out to evaluate the adequacy of the shielding put in place to ensure the safety of the staff, public and oncological patients. The results obtained and measurements made, all fall below the regulatory effective dose limit of 20 mSv per year for staff and 1 mSv per year for the public (author)

  14. Open Rotor Noise Shielding by Blended-Wing-Body Aircraft

    Science.gov (United States)

    Guo, Yueping; Czech, Michael J.; Thomas, Russell H.

    2015-01-01

    This paper presents an analysis of open rotor noise shielding by Blended Wing Body (BWB) aircraft by using model scale test data acquired in the Boeing Low Speed Aeroacoustic Facility (LSAF) with a legacy F7/A7 rotor model and a simplified BWB platform. The objective of the analysis is the understanding of the shielding features of the BWB and the method of application of the shielding data for noise studies of BWB aircraft with open rotor propulsion. By studying the directivity patterns of individual tones, it is shown that though the tonal energy distribution and the spectral content of the wind tunnel test model, and thus its total noise, may differ from those of more advanced rotor designs, the individual tones follow directivity patterns that characterize far field radiations of modern open rotors, ensuring the validity of the use of this shielding data. Thus, open rotor tonal noise shielding should be categorized into front rotor tones, aft rotor tones and interaction tones, not only because of the different directivities of the three groups of tones, but also due to the differences in their source locations and coherence features, which make the respective shielding characteristics of the three groups of tones distinctly different from each other. To reveal the parametric trends of the BWB shielding effects, results are presented with variations in frequency, far field emission angle, rotor operational condition, engine installation geometry, and local airframe features. These results prepare the way for the development of parametric models for the shielding effects in prediction tools.

  15. Calculation of the neutrons shielding in cyclotron accelerator

    International Nuclear Information System (INIS)

    Ribeiro, Martha S.; Sanches, Matias P.; Rodrigues, Demerval L.

    2000-01-01

    The objective of radioprotection in cyclotron facilities is to reduce the dose levels in the workplaces to classify them like supervised areas. In this way, the radiation dose rates in areas occupied by workers during cyclotron operations should not exceed 7,5 μSv/h. In controlled areas these levels are not observed and some rigorous controls must be exerted by administrative procedures or protection mechanisms. The Cyclotron Laboratory at IPEN-CNEN/SP has a cyclotron model Cyclone 30, 30 MeV, used for research and it is also used for radioisotopes production for medical diagnosis and therapeutical applications. Among them, 123 I, 67 Ga and 18 F can be pointed. When accelerator is operating, failures in perforations and paths that conduce to room accelerator can be occur and thus, the dose levels are higher than that established by law. For this reason, a review for shielding structure was necessary in order to optimize radiation dose. The purpose of this work was to determine the shielding thickness and adequate material to diminish the dose rates in workplaces to a value below 7,5 μSv/h. It was used a method to employ the equivalent dose value in the facility areas for neutrons fluency rate for the principal reactions in target irradiation processes. The purposed shielding for the vault doors ensures dose levels lower than established limits to supervised areas. (author)

  16. Production of simplex RNS and ROS by nanosecond pulse N2/O2 plasma jets with homogeneous shielding gas for inducing myeloma cell apoptosis

    Science.gov (United States)

    Liu, Zhijie; Xu, Dehui; Liu, Dingxin; Cui, Qingjie; Cai, Haifeng; Li, Qiaosong; Chen, Hailan; Kong, Michael G.

    2017-05-01

    In this paper, atmospheric pressure N2/O2 plasma jets with homogeneous shielding gas excited by nanosecond pulse are obtained to generate simplex reactive nitrogen species (RNS) and reactive oxygen species (ROS), respectively, for the purpose of studying the simplex RNS and ROS to induce the myeloma cell apoptosis with the same discharge power. The results reveal that the cell death rate by the N2 plasma jet with N2 shielding gas is about two times that of the O2 plasma jet with O2 shielding gas for the equivalent treatment time. By diagnosing the reactive species of ONOO-, H2O2, OH and \\text{O}2- in medium, our findings suggest the cell death rate after plasma jets treatment has a positive correlation with the concentration of ONOO-. Therefore, the ONOO- in medium is thought to play an important role in the process of inducing myeloma cell apoptosis.

  17. 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)

  18. Continuous electrodeionization through electrostatic shielding

    International Nuclear Information System (INIS)

    Dermentzis, Konstantinos

    2008-01-01

    We report a new continuous electrodeionization cell with electrostatically shielded concentrate compartments or electrochemical Faraday cages formed by porous electronically and ionically conductive media, instead of permselective ion exchange membranes. Due to local elimination of the applied electric field within the compartments, they electrostatically retain the incoming ions and act as 'electrostatic ion pumps' or 'ion traps' and therefore concentrate compartments. The porous media are chemically and thermally stable. Electrodeionization or electrodialysis cells containing such concentrate compartments in place of ion exchange membranes can be used to regenerate ion exchange resins and produce deionized water, to purify industrial effluents and desalinate brackish or seawater. The cells can work by polarity reversal without any negative impact to the deionization process. Because the electronically and ionically active media constituting the electrostatically shielded concentrate compartments are not permselective and coions are not repelled but can be swept by the migrating counterions, the cells are not affected by the known membrane associated limitations, such as concentration polarization or scaling and show an increased current efficiency

  19. Current status of JAERI Tokai hot cell facilities

    International Nuclear Information System (INIS)

    Itami, Hiroharu; Morozumi, Minoru; Yamahara, Takeshi

    1992-01-01

    JAERI has 4 hot cell facilities in order to examine high radioactive materials. Three of them, the Research Hot Laboratory, the Reactor Fuel Examination Facility and the Waste Safety Testing Facility are located in the JAERI Tokai site, and the rest is the JMTR Hot Laboratory in the Oarai site. The Research Hot Laboratory (RHL) was constructed for post-irradiation examination (PIE), especially nuclear related basic research experiment, such as metallurgical, chemical and mechanical examination on fuels and materials irradiated in research and test reactors. This facility has 10 large dimension concrete and 38 lead cells. At present the RHL is used for various kinds of examinations of high radioactive samples such as fuels of research and test reactors, power reactors and high temperature testing reactor (HTTR), and structural materials. The Reactor Fuel Examination Facility (RFEF) was designed and constructed for carrying out PIE of irradiated full-size fuel assemblies of light water reactors (LWRs). This facility has a storage pool, 8 concrete and 5 lead cells. They are currently used for safety evaluation on high burnup and advanced lWR fuels as part of the national program. The Waste Safety Testing Facility (WASTEF) was designed and constructed for safety research on long-term storage and disposal of high level radioactive wastes, generated by fuel reprocessing. The WASTEF has 5 concrete cells and 1 lead cell. Examinations on the behavior of various long-lived fission products in a glass form and in a canister and, releasing behavior of them out of a canister are carrying out under the condition at storage. (author)

  20. Evaluation of Shielding Wall Optimization in Lead Slowing Down Spectrometer System

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Ju Young; Kim, Jeong Dong; Lee, Yong Deok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A Lead Slowing Down Spectrometer (LSDS) system is nondestructive technology for analyzing isotope fissile content in spent fuel and pyro processed material, in real time and directly. The high intensity neutron and gamma ray were generated from a nuclear material (Pyro, Spent nuclear fuel), electron beam-target reaction and fission of fissile material. Therefore, shielding analysis of LSDS system should be carried out. In this study, Borax, B{sub 4}C, Li{sub 2}Co{sub 3}, Resin were chosen for shielding analysis. The radiation dose limit (<0.1 μSv/hr) was adopted conservatively at the outer wall surface. The covering could be able to reduce the concrete wall thickness from 5cm to 15cm. The optimized shielding walls evaluation will be used as an important data for future real LSDS facility design and shielding door assessment.

  1. Removal site evaluation report on the bulk shielding facility at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-09-01

    This removal site evaluation report on the Bulk Shielding Facility (BSF) at Oak Ridge National Laboratory was prepared to provide the Environmental Restoration Program with information necessary to evaluate whether hazardous and/or radiological contaminants in and around BSF buildings pose a substantial risk to human health or the environment (i.e., a high probability of adverse effects) and whether remedial site evaluations or removal actions are, therefore, required. A removal site evaluation was conducted at nine areas associated with the BSF. The scope of each evaluation included (1) a search for, and review of, readily available historical records regarding operations and use of the facility (including hazardous substance usage and existing contamination); (2) interviews with facility personnel concerning current and past practices; and (3) a brief walk-through to visually inspect the facility and identify existing hazard areas requiring maintenance actions or remedial evaluation. The results of the removal site evaluation indicate that no substantial risks exist from contaminants present because adequate efforts are being made to contain and control existing contamination and hazardous substances and to protect human health and the environment. At Building 3004, deteriorated and peeling exterior paint has a direct pathway to the storm water drainage system and can potentially impact local surface water during periods of storm water runoff. The paint is assumed to be lead based, thus posing a potential problem. The paint should be sampled and analyzed to determine its lead content and to assess whether a hazard exists. If so, a maintenance action will be necessary to prevent further deterioration and dislodging of the paint. In addition, if the paint contains lead, then a remedial site evaluation should be conducted to determine whether lead from fallen chips has impacted soils in the immediate area of the building

  2. Shielding for a tandem accelerator coupled to linac booster

    International Nuclear Information System (INIS)

    Bhattacharyya, S.; Bisht, J.S.; Venkataraman, G.

    1996-01-01

    Shielding calculation for the Beam-Hall-II of pelletron facility, augmented with linac booster in its phase-II at Nuclear Science Centre, New Delhi, has been done. An estimate is obtained by reduction factor method considering source radiation of monoenergetic neutrons, which is then compared with the detail computation using computer code ALICE considering total energy and angular distribution of neutrons. Another code ASFIT is used to take into account the build up of gamma dose from (n, gamma) reactions within the concrete shield incorporating new radiation weighting factors as recommended by ICRP-60. (author). 8 refs., 2 figs

  3. Use of multiple radiographic techniques for nuclear shielding development

    Energy Technology Data Exchange (ETDEWEB)

    Day, S., E-mail: dayse@mcmaster.ca [McMaster Univ., McMaster Nuclear Reactor, Hamilton, Ontario (Canada)

    2016-01-15

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  4. Use of multiple radiographic techniques for nuclear shielding development

    International Nuclear Information System (INIS)

    Day, S.

    2016-01-01

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  5. Re-evaluation of the shielding adequacy of the brachytherapy treatment room at Korle-Bu teaching hospital, Ghana

    International Nuclear Information System (INIS)

    Arwui, C. C.

    2009-06-01

    Staff and the general public's safety during the operation of the 137 Cs brachytherapy unit at the Korle Bu teaching hospital depends on the adequacy of the shielding of the facility. Shielding design of the brachytherapy unit at the hospital was based on postulated workload and postulated occupancy factors to critical locations at the facility where the public and staff may occupy. This facility has been in existence for the past twelve (12) years and has accumulated operational workload data which differs from the postulated one. A study was carried out to re-evaluate the integrity of the biological shielding of the 137 Cs brachytherapy unit. This study analyzed the accumulated workload data and used the information to perform shielding calculations to verify the adequacy of the biological shielding thicknesses to provide sufficient protection of staff and the public. Dose rate calculations were verified by measurements with calibrated dose rate meters. This provided the basis for determining the current state of protection and safety for staff and the general public. The results show that despite the variation in actual and postulated workloads, the dose rates were below the reference values of 0.5μSv/h for public areas and 7.5μSv/h for controlled areas. It was confirmed that the present shielding thickness of 535 mm can accommodate a high dose rate (HDR) 192 Ir source with activity in the range 370 - 570 GBq with an operational workload of 30 patients per week and an average treatment time of 10 minutes.

  6. Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

    International Nuclear Information System (INIS)

    Margeanu, C.A.

    2011-01-01

    Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

  7. Tough graphene-polymer microcellular foams for electromagnetic interference shielding.

    Science.gov (United States)

    Zhang, Hao-Bin; Yan, Qing; Zheng, Wen-Ge; He, Zhixian; Yu, Zhong-Zhen

    2011-03-01

    Functional polymethylmethacrylate (PMMA)/graphene nanocomposite microcellular foams were prepared by blending of PMMA with graphene sheets followed by foaming with subcritical CO(2) as an environmentally benign foaming agent. The addition of graphene sheets endows the insulating PMMA foams with high electrical conductivity and improved electromagnetic interference (EMI) shielding efficiency with microwave absorption as the dominant EMI shielding mechanism. Interestingly, because of the presence of the numerous microcellular cells, the graphene-PMMA foam exhibits greatly improved ductility and tensile toughness compared to its bulk counterpart. This work provides a promising methodology to fabricate tough and lightweight graphene-PMMA nanocomposite microcellular foams with superior electrical and EMI shielding properties by simultaneously combining the functionality and reinforcement of the graphene sheets and the toughening effect of the microcellular cells.

  8. 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

  9. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Bordier, Gilles; Warin, Dominique; Masson, Michel [CEA/Marcoule Direction, BP 17171 - 30207 - Bagnols-sur-Ceze Cedex (France)

    2008-07-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX{sup TM} process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO{sub 2} powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  10. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    International Nuclear Information System (INIS)

    Bordier, Gilles; Warin, Dominique; Masson, Michel

    2008-01-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX TM process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO 2 powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  11. 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)

  12. Verification of effectiveness of borated water shield for a cyclotron type self-shielded; Verificacao da eficacia da blindagem de agua borada construida para um acelerador ciclotron do tipo autoblindado

    Energy Technology Data Exchange (ETDEWEB)

    Videira, Heber S.; Burkhardt, Guilherme M.; Santos, Ronielly S., E-mail: heber@cyclopet.com.br [Cyclopet Radiofarmacos Ltda., Curitiba, PR (Brazil); Passaro, Bruno M.; Gonzalez, Julia A.; Santos, Josefina; Guimaraes, Maria I.C.C. [Universidade de Sao Paulo (HCFMRP/USP), Sao Paulo, SP (Brazil). Faculdade de Medicina. Hospital das Clinicas; Lenzi, Marcelo K. [Universidade Federal do Parana (UFPR), Curitina (Brazil). Programa de Pos-Graduacao em Engenharia Quimica

    2013-04-15

    The technological advances in positron emission tomography (PET) in conventional clinic imaging have led to a steady increase in the number of cyclotrons worldwide. Most of these cyclotrons are being used to produce {sup 18}F-FDG, either for themselves as for the distribution to other centers that have PET. For there to be safety in radiological facilities, the cyclotron intended for medical purposes can be classified in category I and category II, ie, self-shielded or non-shielded (bunker). Therefore, the aim of this work is to verify the effectiveness of borated water shield built for a cyclotron accelerator-type Self-shielded PETtrace 860. Mixtures of water borated occurred in accordance with the manufacturer’s specifications, as well as the results of the radiometric survey in the vicinity of the self-shielding of the cyclotron in the conditions established by the manufacturer showed that radiation levels were below the limits. (author)

  13. MARS14 deep-penetration calculation for the ISIS target station shielding

    International Nuclear Information System (INIS)

    Nakao, Noriaki; Nunomiya, Tomoya; Iwase, Hiroshi; Nakamura, Takashi

    2004-01-01

    The calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility of Rutherford Appleton Laboratory. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation, a three-dimensional multi-layer technique and energy cut-off method were used considering a spatial statistical balance. Finally, the energy spectra of neutrons behind the very thick shield could be calculated down to the thermal energy with good statistics, and the calculated results typically agree well within a factor of two with the experimental data over a broad energy range. The 12 C(n,2n) 11 C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem

  14. Design and demonstration of adiabatic quantum-flux-parametron logic circuits with superconductor magnetic shields

    International Nuclear Information System (INIS)

    Inoue, Kenta; Narama, Tatsuya; Yamanashi, Yuki; Yoshikawa, Nobuyuki; Takeuchi, Naoki

    2015-01-01

    Adiabatic quantum-flux-parametron (AQFP) logic is an energy-efficient superconductor logic with zero static power and very small dynamic power due to adiabatic switching operations. In order to build large-scale digital circuits, we built AQFP logic cells using superconductor magnetic shields, which are necessary in order to avoid unwanted magnetic couplings between the cells and excitation currents. In preliminary experimental tests, we confirmed that the unwanted coupling became negligibly small thanks to the superconductor shields. As a demonstration, we designed a four-to-one multiplexor and a 16-junction full adder using the shielded logic cells. In both circuits, we confirmed correct logic operations with wide operation margins of excitation currents. These results indicate that large-scale AQFP digital circuits can be realized using the shielded logic cells. (paper)

  15. Specification of the near-Earth space environment with SHIELDS

    International Nuclear Information System (INIS)

    Jordanova, Vania Koleva; Delzanno, Gian Luca; Henderson, Michael Gerard; Godinez, Humberto C.; Jeffery, Christopher Andrew Munn

    2017-01-01

    Here, predicting variations in the near-Earth space environment that can lead to spacecraft damage and failure is one example of “space weather” and a big space physics challenge. A project recently funded through the Los Alamos National Laboratory (LANL) Directed Research and Development (LDRD) program aims at developing a new capability to understand, model, and predict Space Hazards Induced near Earth by Large Dynamic Storms, the SHIELDS framework. The project goals are to understand the dynamics of the surface charging environment (SCE), the hot (keV) electrons representing the source and seed populations for the radiation belts, on both macro- and micro-scale. Important physics questions related to particle injection and acceleration associated with magnetospheric storms and substorms, as well as plasma waves, are investigated. These challenging problems are addressed using a team of world-class experts in the fields of space science and computational plasma physics, and state-of-the-art models and computational facilities. A full two-way coupling of physics-based models across multiple scales, including a global MHD (BATS-R-US) embedding a particle-in-cell (iPIC3D) and an inner magnetosphere (RAM-SCB) codes, is achieved. New data assimilation techniques employing in situ satellite data are developed; these provide an order of magnitude improvement in the accuracy in the simulation of the SCE. SHIELDS also includes a post-processing tool designed to calculate the surface charging for specific spacecraft geometry using the Curvilinear Particle-In-Cell (CPIC) code that can be used for reanalysis of satellite failures or for satellite design.

  16. A novel comprehensive utilization of vanadium slag: As gamma ray shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Mengge [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Xue, Xiangxin, E-mail: xuexx@mail.neu.edu.cn [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Yang, He; Liu, Dong [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Wang, Chao [Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, Zhefu [Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-11-15

    Highlights: • A novel comprehensive utilization method for vanadium slag is proposed. • Shielding properties of vanadium slag are better than ordinary concrete. • HVL of vanadium slag is between Lead and concrete to shield {sup 60}Co gamma ray. • HVL of composite is higher than concrete when adding amount of vanadium slag is 900. • Composite can be used as injecting mortar for cracks developed in concrete shields. - Abstract: New exploration of vanadium slag as gamma ray shielding material was proposed, the shielding properties of vanadium slag was higher than concrete when the energy of photons was in 0.0001 MeV–100000 MeV. Vanadium slag/epoxy resin composites were prepared, shielding and material properties of materials were tested by {sup 60}Co gamma ray, simultaneous DSC-TGA, electronic universal testing machine and scanning electron microscopy, respectively. The results showed that the shielding properties of composite would be better with the increase of vanadium slag addition amount. The HVL (half value layer thickness) of vanadium slag was between Lead and concrete while composite was higher than concrete when the addition amount of vanadium slag was 900 used as material to shield {sup 60}Co gamma ray, also the resistance temperature of composite was about 215 °C and the bending strength was over 10 MPa. The composites could be used as injecting mortar for cracks developed in biological concrete shields, coating for the floor of the nuclear facilities, and shielding materials by itself.

  17. Building of a facility for the handling of kilo-curie amounts of gamma emitters

    International Nuclear Information System (INIS)

    Germond, Ph.

    1960-01-01

    A hot cell designed to handle up to 1000 curies of cobalt-60 has been built in a preexisting shielded room, in order to make optimum use of available space. Heavy containers can be rolled in or out of the cell. Handling performed with two manipulators designed and made by French manufacturers, one of them is pneumatically operated and the other one is mechanical. The general shape of the facility is that of an L. (author) [fr

  18. Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

    Directory of Open Access Journals (Sweden)

    Ahmed Amin El Said Abd El Hameed

    2015-08-01

    Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

  19. Development of remote handling technology for nuclear fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Sakai, Akira; Miura, Noriaki; Kozaka, Tetsuo; Hamada, Takashi

    2015-01-01

    Remote handling technology has been systematically developed for nuclear fuel cycle facilities in Japan since 1970s, primarily in parallel with the development of reprocessing and HLLW (High Level Liquid Waste) vitrification process. In case of reprocessing and vitrification process to handle highly radioactive and hazardous materials, the most of components are installed in the radiation shielded hot cells and operators are not allowed to enter the work area in the cells for operation and maintenance. Therefore, a completely remote handling system is adopted for the cells to reduce radiation doses of operators and increase the availability of the facility. The hot cells are generally designed considering the scale of components (laboratory, demonstration, or full-scale), the function of the systems (chemical process, material handling, dismantling, decontamination, or chemical analysis), and the environmental conditions (radiation dose rate, airborne concentration, surface contamination, or fume/mist/dust). Throughout our domestic development work for remote handling technology, the concept of the large scale integrated cell has been adopted rather than a number of small scale separated cells, for the reasons to reduce the total installation space and the number of remote handling equipment required for the each cell as much as possible. In our domestic remote maintenance design, several new concepts have been developed, tested, and demonstrated in the Tokai Virtrification Facility (TVF) and the Rokkasho HLLW Vitrification and Storage Facility (K-facility). Layout in the hot cells, the performance of remote handling equipment, and the structure of the in-cell components are important factors for remote maintenance design. In case of TVF (hot tests started in 1995), piping and vessels are prefabricated in the rack modules and installed in two lines on both sides of the cell. These modules are designed to be remotely replaced in the whole rack. Two overhead cranes

  20. Construction of concrete hot cells

    International Nuclear Information System (INIS)

    1981-12-01

    The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design hasis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

  1. Construction of concrete hot cells

    International Nuclear Information System (INIS)

    1980-09-01

    The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design basis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

  2. Radiation shielding and health physics instrumentation for PET medical cyclotrons

    International Nuclear Information System (INIS)

    Mukherjee, B.

    2002-01-01

    Full text: Modern Medical Cyclotrons produce a variety of short-lived positron emitting PET radioisotopes, and as a result are the source of intense neutron and gamma radiations. Since such cyclotrons are housed within hospitals or medical clinics, there is significant potential for un-intentional exposure to staff or patients in proximity to cyclotron facilities. Consequently, the radiological hazards associated with Cyclotrons provide the impetus for an effective radiological shielding and continuous monitoring of various radiation levels in the cyclotron environment. Management of radiological hazards is of paramount importance for the safe operation of a Medical Cyclotron facility. This work summarised the methods of shielding calculations for a compact hospital based Medical Cyclotron currently operating in Canada, USA and Australia. The design principle and operational history of a real-time health physics monitoring system (Watchdog) operating at a large multi-energy Medical Cyclotron is also highlighted

  3. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hyun Whee; Lee, Kang Moo; Koo, Jun Mo; Jung, In Ha; Lee, Jong Ryeul; Kim, Sung Whan; Bae, Sang Min; Cho, Kang Whon; Sung, Suk Jong

    1989-02-01

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  4. Military Handbook. Grounding, Bonding, and Shielding for Electronic Equipments and Facilities. Volume 1. Basic Theory

    Science.gov (United States)

    1987-12-29

    when the air or gas stream contains particulate matter. b. Pulverized materials passing through chutes or pneumatic conveyors . c. Nonconductive power...Hanover NH, 1971, AD 722 221. 146.Oakley, R.J., "Surface Transfer Impedance and Cable Shielding Design ," Wire Journal, Vol 4, No. 3, March 1971, pp...including considerations of grounding, bonding, and shielding in all phases of design , construction, operation, and maintenance of electronic equipment

  5. One Year assessment of shielding for a multi-energy linear accelerator

    International Nuclear Information System (INIS)

    Lee, Jae Gi; Carlson, Joel; Lee, Hyun Seok; Ye, Sung Joon; Chung, Jin Beom; Kim, Jae Sung; Kim, Jung In

    2014-01-01

    In 2005, the publication of Report No. 151 of the National Council on Radiation Protection and Measurements (NCRP) suggested shielding methodologies along with shielding data. Recently, intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) have become more widely used for cancer treatment. Thus, we analyzed shielding parameters for a multi-energy medical linear accelerator using the VMAT technique. Calculated total workload was similar to the recommendation of NCRP Report No. 49 and No. 51. However, these results were higher than the previous results in the NCRP Report No. 151. Also, the VMAT technique uses an intensity modulated beams with various gantry angles so that scattered and leakage doses should be carefully considered by retrospective analysis using the treatment data from each facility

  6. One Year assessment of shielding for a multi-energy linear accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Gi; Carlson, Joel; Lee, Hyun Seok; Ye, Sung Joon [Seoul National University Graduate School of Convergence Science and Technology, Seoul (Korea, Republic of); Chung, Jin Beom; Kim, Jae Sung [Dept. of Radiation Oncology, Seoul National University Bundang Hospital, Seoul (Korea, Republic of); Kim, Jung In [Dept. of of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of)

    2014-11-15

    In 2005, the publication of Report No. 151 of the National Council on Radiation Protection and Measurements (NCRP) suggested shielding methodologies along with shielding data. Recently, intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) have become more widely used for cancer treatment. Thus, we analyzed shielding parameters for a multi-energy medical linear accelerator using the VMAT technique. Calculated total workload was similar to the recommendation of NCRP Report No. 49 and No. 51. However, these results were higher than the previous results in the NCRP Report No. 151. Also, the VMAT technique uses an intensity modulated beams with various gantry angles so that scattered and leakage doses should be carefully considered by retrospective analysis using the treatment data from each facility.

  7. Assessment of shielding integrity of Co-60 gamma irradiation-ray scanner at Aflao Border, Ghana

    International Nuclear Information System (INIS)

    Agbemafo, Edwin Capacity

    2016-07-01

    This study examines the current state of the shielding integrity of the 38.7 TBq Co-60 gamma ray scanner with an average energy of 1.25 MeV operated by NICK TC Scan Limited, which has been in use for destination inspection at Aflao Border of Ghana, for the past six years, (2010-2016). The facility uses a high energy ionizing radiation in its operation; therefore continuous adequacy of the installed biological shielding is critical to the protection and safety of the workers and the general public. The workload of the facility has increased since its commissioning, requiring the review of the status of the installed shielding. Theoretical calculations for dose rates and barrier thicknesses based on tenth – value layer (TVL) concept and NCRP 151, 2005 recommendations, were done around the scanning facility using the current operational data. The results were then compared with the measured dose rates and the shielding thickness constituted during the commissioning stage, and international standards. Calculated dose rate at commissioning state ranges from 0.6μSv/hr to 2.4μSv/hr with an average dose rate of 1.43μSv/hr and that of the current operational state ranges from 1.1μSv/hr to 2.6μSv/hr with an average dose rate of 1.54μSv/hr, indicating an increase of 7.9%. Even though the dose rates were all below the recommended dose limit of 20μSvh"-"1 by NCRP, there has been an increase in dose to the staff and the general public. It has been observed that, the workload has increased three-fold from the commissioning stage to current operational state over the past six years. The assessment done on the installed shielding using the current operational data indicates that the shielding is inadequate in providing protection for the general public and the workers against X-ray radiation source of energy of at least 6MeV, and therefore the facility in its current state cannot be used to house a linear accelerator of energy up to 10MeV. (au)

  8. Omni-directional selective shielding material based on amorphous glass coated microwires.

    Science.gov (United States)

    Ababei, G; Chiriac, H; David, V; Dafinescu, V; Nica, I

    2012-01-01

    The shielding effectiveness of the omni-directional selective shielding material based on CoFe-glass coated amorphous wires in 0.8 GHz-3 GHz microwave frequency range is investigated. The measurements were done in a controlled medium using a TEM cell and in the free space using horn antennas, respectively. Experimental results indicate that the composite shielding material can be developed with desired shielding effectiveness and selective absorption of the microwave frequency range by controlling the number of the layers and the length of microwires.

  9. Study on the bearing capacity of embedded chute on shield tunnel segment

    Science.gov (United States)

    Fanzhen, Zhang; Jie, Bu; Zhibo, Su; Qigao, Hu

    2018-05-01

    The method of perforation and steel implantation is often used to fix and install pipeline, cables and other facilities in the shield tunnel, which would inevitably do damage to the precast segments. In order to reduce the damage and the resulting safety and durability problems, embedded chute was set at the equipment installation in one shield tunnel. Finite element models of segment concrete and steel are established in this paper. When water-soil pressure calculated separately and calculated together, the mechanical property of segment is studied. The bearing capacity and deformation of segment are analysed before and after embedding the chute. Research results provide a reference for similar shield tunnel segment engineering.

  10. Proposed Californium-252 User Facility for Neutron Science at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Martin, R.C.; Laxson, R.R.; Knauer, J.B.

    1996-01-01

    The Radiochemical Engineering Development Center (REDC) at ORNL has petitioned to establish a Californium-252 User Facility for Neutron Science for academic, industrial, and governmental researchers. The REDC Californium Facility (CF) stores the national inventory of sealed 252 Cf neutron source for university and research loans. Within the CF, the 252 Cf storage pool and two uncontaminated hot cells currently in service for the Californium Program will form the physical basis for the User Facility. Relevant applications include dosimetry and experiments for neutron tumor therapy; fast and thermal neutron activation analysis of materials; experimental configurations for prompt gamma neutron activation analysis; neutron shielding and material damage studies; and hardness testing of radiation detectors, cameras, and electronics. A formal User Facility simplifies working arrangements and agreements between US DOE facilities, academia, and commercial interests

  11. Electromagnetic shielding

    International Nuclear Information System (INIS)

    Tzeng, Wen-Shian V.

    1991-01-01

    Electromagnetic interference (EMI) shielding materials are well known in the art in forms such as gaskets, caulking compounds, adhesives, coatings and the like for a variety of EMI shielding purposes. In the past, where high shielding performance is necessary, EMI shielding has tended to use silver particles or silver coated copper particles dispersed in a resin binder. More recently, aluminum core silver coated particles have been used to reduce costs while maintaining good electrical and physical properties. (author). 8 figs

  12. Spectroscopic Study of Radiation around the Leksell Gamma Knife for Room Shielding Applications

    OpenAIRE

    Hubert, Alexis

    2017-01-01

    Any center planning to install a Gamma Knife radiosurgery unit has to provide for an efficient shielding of the treatment room, to protect the patient, the staff and the public, against undesired radiation. The shielding barrier design is controlled by national and international recommendations; the reference documents for gamma ray radiotherapy facilities are the National Council on Radiation Protection and Measurements (NCRP) reports 49 and 151. However, some facts highlighted in this thesi...

  13. 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.

  14. About the Scythian Shields

    Directory of Open Access Journals (Sweden)

    About the Scythian Shields

    2017-10-01

    Full Text Available Shields played major role in the armament system of the Scythians. Made from organic materials, they are poorly traced on the materials of archaeological excavations. Besides, scaly surface of shields was often perceived in practice as the remnants of the scaly armor. E. V. Chernenko was able to discern the difference between shields’ scaly plates and armor scales. The top edge of the scales was bent inwards, and shield plates had a wire fixation. These observations let significantly increase the number of shields, found in the burial complexes of the Scythians. The comparison of archaeological materials and the images of Scythian warriors allow distinguishing the main forms of Scythian shields. All shields are divided into fencing shields and cover shields. The fencing shields include round wooden shields, reinforced with bronze sheet, and round moon-shaped shields with a notch at the top, with a metal scaly surface. They came to the Scythians under the Greek influence and are known in the monuments of the 4th century BC. Oval shields with scaly surface (back cover shields were used by the Scythian cavalry. They protected the rider in case of frontal attack, and moved back in case of maneuver or closein fighting. Scythian battle tactics were based on rapid approaching the enemy and throwing spears and further rapid withdrawal. Spears stuck in the shields of enemies, forcing them to drop the shields, uncover, and in this stage of the battle the archers attacked the disorganized ranks of the enemy. That was followed by the stage of close fight. Oval form of a wooden shield with leather covering was used by the Scythian infantry and spearmen. Rectangular shields, including wooden shields and the shields pleached from rods, represented a special category. The top of such shield was made of wood, and a pleached pad on leather basis was attached to it. This shield could be a reliable protection from arrows, but it could not protect against javelins

  15. Survivor shielding. Part C. Improvements in terrain shielding

    International Nuclear Information System (INIS)

    Egbert, Stephen D.; Kaul, Dean C.; Roberts, James A.; Kerr, George D.

    2005-01-01

    A number of atomic-bomb survivors were affected by shielding provided by terrain features. These terrain features can be a small hill, affecting one or two houses, or a high mountain that shields large neighborhoods. In the survivor dosimetry system, terrain shielding can be described by a transmission factor (TF), which is the ratio between the dose with and without the terrain present. The terrain TF typically ranges between 0.1 and 1.0. After DS86 was implemented at RERF, the terrain shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, an improvement in the terrain model was implemented at RERF in the revised DS86 code, but the documentation was not published. It is now presented in this section. The solution to the terrain shielding in front of a house is described in this section. The problem of terrain shielding of survivors behind Hijiyama mountain at Hiroshima and Konpirasan mountain at Nagasaki has also been recognized, and a solution to this problem has been included in DS02. (author)

  16. Neutron shielding and activation of the MASTU device and surrounds

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, David, E-mail: david.taylor@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lilley, Steven; Turner, Andrew [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, Andrew [Now at College of Engineering, University of Wisconsin, Madison, WI 53706 (United States)

    2014-10-15

    Highlights: We model neutron shielding for the planned MASTU device; nadequacies in the existing shielding design are remedied; Levels of public exposure are considered; We model activated gamma emission for the device under a worst case scenario. Abstract: A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day. The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.

  17. Operation of post-irradiation examination facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, E. G.; Jeon, Y. B.; Ku, D. S.

    1996-12-01

    In 1996, the post-irradiation examination(PIE) of nuclear fuels was performed as follows. It has been searched for the caution of defection of defected fuel rods of Youngkwang-4 reactor through NDT and metallographic examination that had been required by KEPCO. And in-pool inspection of Kori-1 spent fuel assembly(FO2) was carried out. HVAC system and pool water treatment system have been operated to maintain the facility safely, and electric power supply system was checked and maintained for the normal and steady supply electric power to the facility. Image processing software was developed for measurement of defection of spent fuel rods. Besides, a radiation shielding glove box was fabricated and a hot cell compressor for volume reduction of radioactive materials was fabricated and installed in hot cell. Safeguards of nuclear materials were implemented in strict accordance with the relevant Korean rules and regulations as well as the international non-proliferation regime. Also the IAEA inspection was carried out on the quarterly basis. (author). 31 tabs., 71 figs., 4 refs.

  18. Decision-making methodology of optimal shielding materials by using fuzzy linear programming

    International Nuclear Information System (INIS)

    Kanai, Y.; Miura, T.; Hirao, Y.

    2000-01-01

    The main purpose of our studies are to select materials and determine the ratio of constituent materials as the first stage of optimum shielding design to suit the individual requirements of nuclear reactors, reprocessing facilities, casks for shipping spent fuel, etc. The parameters of the shield optimization are cost, space, weight and some shielding properties such as activation rates or individual irradiation and cooling time, and total dose rate for neutrons (including secondary gamma ray) and for primary gamma ray. Using conventional two-valued logic (i.e. crisp) approaches, huge combination calculations are needed to identify suitable materials for optimum shielding design. Also, re-computation is required for minor changes, as the approach does not react sensitively to the computation result. Present approach using a fuzzy linear programming method is much of the decision-making toward the satisfying solution might take place in fuzzy environment. And it can quickly and easily provide a guiding principle of optimal selection of shielding materials under the above-mentioned conditions. The possibility or reducing radiation effects by optimizing the ratio of constituent materials is investigated. (author)

  19. Design, fabrication, installation and shielding integrity testing of source storage container for automatic source movement system used in TLD calibration facility

    International Nuclear Information System (INIS)

    Subramanian, V.; Baskar, S.; Annalakshmi, O.; Jose, M.T.; Jayshree, C.P.; Choudry, Shreelatha

    2012-01-01

    A state-of-art TLD laboratory has been commissioned in January 2000 at Radiological Safety Division of Indira Gandhi Centre for Atomic Research (IGCAR). The laboratory provides personnel monitoring service to 2000 occupational workers from Indira Gandhi Centre for Atomic Research and Bhabha Atomic Research Centre facilities. The laboratory has been accredited by the Radiation Safety Systems Division (RSSD), Bhabha Atomic Research Centre (BARC) since year 2002. The laboratory has exclusive facility for the calibration of the TLD cards. As apart of accreditation procedure and taking into account of geometry effect, the dose rate at the card position is determined by the accreditation authorities by using graphite chamber (secondary or national standard instrument) and often re estimated by a condenser R meter (M/s Victoreen, Germany) by our laboratory. As per the regulatory requirement, the exposure protocols should be automated. Towards this an automatic source movement system has been augmented in the calibration facility. By using the system, the source will be brought to the irradiation position by pneumatically and exposures will be terminated by counter, timer and triggering system. To accomplish this task a lead container has been designed, fabricated and mounted at the beneath of the calibration table for the storage of source. As per the automation process, a lead container for the source storage has been designed and installed beneath to the Calibration Table. The container was designed to hold a 3Ci 137 Cs source, but present activity of the source is 1.2Ci. Hence, the shielding integrity was tested with higher active source (1.7Ci 60 Co). The dose rate measured outside on the circumference of the container at the middle of the source is found to be the same as calculated using QAD CGGP calculations. The top plug is so designed to avoid inadvertent upward movement of the source. Though, the shielding was not adequate on top of the top plug, however it does

  20. Discussion on the Standardization of Shielding Materials — Sensitivity Analysis of Material Compositions

    Directory of Open Access Journals (Sweden)

    Ogata Tomohiro

    2017-01-01

    Full Text Available The overview of standardization activities for shielding materials is described. We propose a basic approach for standardizing material composition used in radiation shielding design for nuclear and accelerator facilities. We have collected concrete composition data from actual concrete samples to organize a representative composition and its variance data. Then the sensitivity analysis of the composition variance has been performed through a simple 1-D dose calculation. Recent findings from the analysis are summarized.

  1. Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J. M.; Crawford, C. L.

    2013-04-26

    The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

  2. Development of point Kernel radiation shielding analysis computer program implementing recent nuclear data and graphic user interfaces

    International Nuclear Information System (INIS)

    Kang, S.; Lee, S.; Chung, C.

    2002-01-01

    There is an increasing demand for safe and efficient use of radiation and radioactive work activity along with shielding analysis as a result the number of nuclear and conventional facilities using radiation or radioisotope rises. Most Korean industries and research institutes including Korea Power Engineering Company (KOPEC) have been using foreign computer programs for radiation shielding analysis. Korean nuclear regulations have introduced new laws regarding the dose limits and radiological guides as prescribed in the ICRP 60. Thus, the radiation facilities should be designed and operated to comply with these new regulations. In addition, the previous point kernel shielding computer code utilizes antiquated nuclear data (mass attenuation coefficient, buildup factor, etc) which were developed in 1950∼1960. Subsequently, the various nuclear data such mass attenuation coefficient, buildup factor, etc. have been updated during the past few decades. KOPEC's strategic directive is to become a self-sufficient and independent nuclear design technology company, thus KOPEC decided to develop a new radiation shielding computer program that included the latest regulatory requirements and updated nuclear data. This new code was designed by KOPEC with developmental cooperation with Hanyang University, Department of Nuclear Engineering. VisualShield is designed with a graphical user interface to allow even users unfamiliar to radiation shielding theory to proficiently prepare input data sets and analyzing output results

  3. Shielding benchmark problems

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

    1978-09-01

    Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

  4. The Solid Waste Transfer Pit an FFTF [Fast Flux Test Facility] station for fuel and reflector transfer

    International Nuclear Information System (INIS)

    Milewski, A.J.; Thomson, J.D.

    1985-01-01

    This paper describes the design aspects of a proposed FFTF facility called the Solid Waste Transfer Pit (SWTP). This new facility will be used as a transfer station for reflectors and/or low decay heat fuel assemblies when movement from the Fuel Storage Facility for sodium removal and subsequent reprocessing is to be accomplished. The SWTP utilizes a shielded cell cover with an integral manually-operated gate valve for top loading of a transfer vessel. The cost-effective design discussed herein provides a practical approach using state-of-the-art concepts while assuring safe and reliable operation

  5. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1990-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility, a magnetic mirror device at Oak Ridge National Laboratory, using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages within 50% of those expected in tokamaks. The time-varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with a Langmuir probe. Ion energies were measured with a gridded energy analyser located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurements show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Plasma parameters and ion energies have been correlated with the antenna current and used in a computational model of the plasma sheath to predict the amount of erosion expected from the Faraday shield elements exposed to plasma. Predictions of light ion sputtering of candidate Faraday shield materials are presented

  6. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  7. Shielding design of disposal container for disused sealed radioactive source

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk Hoon; Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of)

    2017-06-15

    Disused Sealed Radioactive Sources (DSRSs), which are stored temporally in the centralized storage facility of Korea Radioactive Waste Agency (KORAD), will be disposed of in the low- and intermediate-level radioactive waste disposal facility located in Wolsong. Accordingly, the future plan on DSRS disposal should be established as soon as possible in connection with the construction and operation plan of disposal facility. In this study, as part of developing the systematic management plan, the radiation shielding analysis for three types of disposal container was performed for all kinds of radionuclides (excluding mixed sources) contained in DSRSs generated from domestic area using MicroShield and MCNP5 codes in consideration of the preliminary post-closure safety assessment result for disposal options, source-specific characteristics, and etc. In accordance with the analysis result, thickness of inner container for general disposal container and dimensions (i.e. diameter and height) of inner capsule for two types of special disposal container were determined as 3 mm, OD40×H120 mm (for type 1), and OD100× H240 mm (for type 2), respectively. These values were reflected in the conceptual design of DSRS disposal container, and the structural integrity of each container was confrmed through the structural analysis carried out separately from this study. Given the shielding and structural analysis results, the conceptual design derived from this study sufficiently fulfills the technical standards in force and the design performance level. And consequently, it is judged that the safe management for DSRSs to be disposed of is achieved by utilizing the disposal container with the conceptual design devised.

  8. Shielding design of disposal container for disused sealed radioactive source

    International Nuclear Information System (INIS)

    Kim, Suk Hoon; Kim, Ju Youl

    2017-01-01

    Disused Sealed Radioactive Sources (DSRSs), which are stored temporally in the centralized storage facility of Korea Radioactive Waste Agency (KORAD), will be disposed of in the low- and intermediate-level radioactive waste disposal facility located in Wolsong. Accordingly, the future plan on DSRS disposal should be established as soon as possible in connection with the construction and operation plan of disposal facility. In this study, as part of developing the systematic management plan, the radiation shielding analysis for three types of disposal container was performed for all kinds of radionuclides (excluding mixed sources) contained in DSRSs generated from domestic area using MicroShield and MCNP5 codes in consideration of the preliminary post-closure safety assessment result for disposal options, source-specific characteristics, and etc. In accordance with the analysis result, thickness of inner container for general disposal container and dimensions (i.e. diameter and height) of inner capsule for two types of special disposal container were determined as 3 mm, OD40×H120 mm (for type 1), and OD100× H240 mm (for type 2), respectively. These values were reflected in the conceptual design of DSRS disposal container, and the structural integrity of each container was confrmed through the structural analysis carried out separately from this study. Given the shielding and structural analysis results, the conceptual design derived from this study sufficiently fulfills the technical standards in force and the design performance level. And consequently, it is judged that the safe management for DSRSs to be disposed of is achieved by utilizing the disposal container with the conceptual design devised

  9. The new MAW scrap processing facility

    International Nuclear Information System (INIS)

    Kueppers, L.

    1994-01-01

    The shielded bunker for heat-generating waste attached to the MAW scrap processing cell will be modified and extended to comprise several MAW scrap processing cells of enhanced throughput capacity, and a new building to serve as an airlock and port for acceptance of large shipping casks (shipping cask airlock, TBS). The new facility is to process scrap from decommissioned nuclear installations, and in addition radwaste accrued at operating plants of utilities. This will allow efficient and steady use of the new MAW scrap processing facility. The planning activities for modification and extension are based on close coordination between KfK and the GNS mbH, in order to put structural dimensioning and capacity planning on a realistic basis in line with expected amounts of radwaste from operating nuclear installations of utilities. The paper indicates the currently available waste amount assessments covering solid radwaste (MAW) from the decommissioning of the WAK, MZFR, and KNK II, and existing waste amounts consisting of core internals of German nuclear power plant. The figures show that the MAW scrap processing facility will have to process an overall bulk of about 1100 Mg of solid waste over the next ten years to come. (orig./HP) [de

  10. 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)

  11. Development of a computational code for calculations of shielding in dental facilities; Desenvolvimento de um codigo computacional para calculos de blindagem em instalacoes odontologicas

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper is prepared in order to address calculations of shielding to minimize the interaction of patients with ionizing radiation and / or personnel. The work includes the use of protection report Radiation in Dental Medicine (NCRP-145 or Radiation Protection in Dentistry), which establishes calculations and standards to be adopted to ensure safety to those who may be exposed to ionizing radiation in dental facilities, according to the dose limits established by CNEN-NN-3.1 standard published in September / 2011. The methodology comprises the use of computer language for processing data provided by that report, and a commercial application used for creating residential projects and decoration. The FORTRAN language was adopted as a method for application to a real case. The result is a programming capable of returning data related to the thickness of material, such as steel, lead, wood, glass, plaster, acrylic, acrylic and leaded glass, which can be used for effective shielding against single or continuous pulse beams. Several variables are used to calculate the thickness of the shield, as: number of films used in the week, film load, use factor, occupational factor, distance between the wall and the source, transmission factor, workload, area definition, beam intensity, intraoral and panoramic exam. Before the application of the methodology is made a validation of results with examples provided by NCRP-145. The calculations redone from the examples provide answers consistent with the report.

  12. National Low-Temperature Neutron-Irradiation Facility

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Klabunde, C.E.; Young, F.W. Jr.

    1983-08-01

    The Materials Sciences Division of the United States Department of Energy will establish a National Low Temperature Neutron Irradiation Facility (NLTNIF) which will utilize the Bulk Shielding Reactor (BSR) located at Oak Ridge National Laboratory. The facility will provide high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. This report describes the planned experimental capabilities of the new facility

  13. Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory East

    International Nuclear Information System (INIS)

    Neimark, L.A.; Jackson, W.D.; Donahue, D.A.

    1979-01-01

    The Alpha-Gamma Hot-Cell Facility has been in operation at Argonne National Laboratory East (ANL-E) for 15 years. The facility was designed for plutonium research in support of ANL's LMFBR program. The facility consists of a kilocurie, nitrogen-atmosphere alpha-gamma hot cell and supporting laboratories. Modifications to the facility and its equipment have been made over the years as the workload and nature of the work changed. These modifications included inerting the entire hot cell, adding four work stations, modifying in-loading procedures and examination equipment to handle longer test articles, and changing to a new sodium-vapor lighting system. Future upgrading includes the addition of a decontamination and repair facility, use of radio-controlled transfer carts, refurbishment of the zinc bromide windows, and the installation of an Auger microprobe

  14. 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.

  15. Electromagnetic shielding formulae

    International Nuclear Information System (INIS)

    Dahlberg, E.

    1979-02-01

    This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

  16. Neutron transmission benchmark problems for iron and concrete shields in low, intermediate and high energy proton accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakane, Yoshihiro; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Katsumi [and others

    1996-09-01

    Benchmark problems were prepared for evaluating the calculation codes and the nuclear data for accelerator shielding design by the Accelerator Shielding Working Group of the Research Committee on Reactor Physics in JAERI. Four benchmark problems: transmission of quasi-monoenergetic neutrons generated by 43 MeV and 68 MeV protons through iron and concrete shields at TIARA of JAERI, neutron fluxes in and around an iron beam stop irradiated by 500 MeV protons at KEK, reaction rate distributions inside a thick concrete shield irradiated by 6.2 GeV protons at LBL, and neutron and hadron fluxes inside an iron beam stop irradiated by 24 GeV protons at CERN are compiled in this document. Calculational configurations and neutron reaction cross section data up to 500 MeV are provided. (author)

  17. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

    International Nuclear Information System (INIS)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Boeck, Helmuth; Steinhauser, Georg

    2011-01-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10 9 cm -2 s -1 at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: → Neutron activation is an important process for the waste management of nuclear facilities. → Biological shield of the TRIGA reactor Vienna has been topic of investigation. → Flux values allow a categorization of the concrete concerning radiation protection legislation. → Reactor installations are of great importance as neutron sources into the biological shield. → Every installation shows distinguishable flux profiles.

  18. Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

    International Nuclear Information System (INIS)

    Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei

    2012-01-01

    Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

  19. Conceptual layout design of CFETR Hot Cell Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Zheng, E-mail: gongz@mail.ustc.edu.cn [University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Qi, Minzhong, E-mail: qiminzhong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Cheng, Yong, E-mail: chengyong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • This article proposed a conceptual layout design for CFETR. • The design principles are to support efficient maintenance to ensure the realization of high duty time. • The preliminary maintenance process and logistics are described in detail. • Life cycle management, maneuverability, risk and safety are in the consideration of design. - Abstract: CFETR (China Fusion Engineering Test Reactor) is new generation of Tokomak device beyond EAST in China. An overview of hot cell layout design for CFETR has been proposed by ASIPP&USTC. Hot Cell, as major auxiliary facility, not only plays a pivotal role in supporting maintenance to meet the requirements of high duty time 0.3–0.5 but also supports installation and decommissioning. Almost all of the Tokomak devices are lateral handling internal components like ITER and JET, but CFETR maintain the blanket module from 4 vertical ports, which is quite a big challenge for the hot cell layout design. The activated in-vessel components and several diagnosis instruments will be repaired and refurbished in the Hot Cell Facility, so the appropriate layout is very important to the Hot Cell Facility to ensure the high duty time, it is divided into different parts equipped with a variety of RH equipment and diagnosis devices based on the functional requirements. The layout of the Hot Cell Facility should make maintenance process more efficient and reliable, and easy to service and rescue when a sudden events taking place, that is the capital importance issue considered in design.

  20. Shielding member for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori

    1997-06-30

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

  1. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  2. New facilities of the ECN hot cell laboratory

    International Nuclear Information System (INIS)

    Duijves, K.A.; Konings, R.J.M.

    1996-04-01

    A description is given of two recent expansions of the ECN Hot Cell Laboratory in Petten; a production facility for molybdenum-99 and an actinide laboratory, a special facility to investigate unirradiated alpha- and beta-active samples. (orig.)

  3. Radiation shielding provided by residential houses in Japan in reactor accidents accompanied with atmospheric release

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro; Minami, Kentaro

    1991-01-01

    The present report describes the radiation shielding effect of houses in Japan against the radioactive cloud resulting from a major reactor accident accompanied with atmospheric release. The shielding factor of houses, the ratio of indoor exposure rate to outdoor one, has been studied for the semi-infinite and finite clouds which contain γ-emitting radionuclides released from a reactor facility. The shielding factor of houses against γ-rays from the radioactive cloud decreases gradually with release delay time and keeps a minimum during the period from 50 to 1000 hours after reactor shutdown while 133 Xe predominates in the cloud. Radioiodines mixed in the cloud raise slightly the shielding factor, and the factor depends little on the shape of the cloud. A set of shielding factors for the use of emergency planning was consequently proposed as 0.4 for simple ferroconcrete residential house and 0.9 for other ordinary ones. (author)

  4. Drop test facility available to private industry

    International Nuclear Information System (INIS)

    Shappert, L.B.; Box, W.D.

    1983-01-01

    In 1978, a virtually unyielding drop test impact pad was constructed at Oak Ridge National Laboratory's (ORNL's) Tower Shielding Facility (TSF) for the testing of heavy shipping containers designed for transporting radioactive materials. Because of the facility's unique capability for drop-testing large, massive shipping packages, it has been identified as a facility which can be made available for non-DOE users

  5. Evaluation of a pyrex glass shield for the dose reduction in extremities to manipulate a 90 Sr- 90 Y generator

    International Nuclear Information System (INIS)

    Ayra P, F.E.; Xiques C, A.; Torres B, M.B.

    2006-01-01

    The production of Y-90 of high activity it specifies (free of payee) for their use in radioimmunotherapy uses the Strontium 90 as isotope source. Depending on the method employee for the separation of both isotopes several types of generators are described in different bibliographies. The column generator used in the facilities of the Center of Isotopes requires of a frequent manipulation causing significant dose in the skin of the extremities due to the exhibition to the radiation beta of high energy. The properties of the shieldings for this radiation type have been well studied Y they consist in several publications. To be in correspondence with requirements of radiological protection in the Cuban legislation, the column was covered with a tube of glass pyrex of 5 mm of thickness and it was monitored the exposure with an ionization chamber. At the own time, the shielding using the Monte Carlo method was evaluated. It was used the MCNP 4C code to simulate the absorption of the beta particles generated in the process of disintegration of the Sr-90 and Y-90 in the glass shielding. The column generator and the fluence of beta particles were modeled in different points inside the shielding to determine if the experimentally measured values correspond to electrons that were not absorbed or to the weak stopping radiation generated in the glass due to the deceleration of these particles. A cylinder of 4 mm of diameter simulates the source (it dilutes) and a tube of walls of 6 mm of thickness simulates the shielding more the wall of the column around the generator. This it was divided in cells of 1 mm of thickness and the energy deposited in them was evaluated. The results show that all the electrons generated in the source are absorbed in the shielding and the exposure rates decrease in more of 78 times using the 5 mm of pyrex glass. The doses in extremities to the operators of the generator don't surpass the 70 mSv by year that is the dose restriction imposed in the

  6. Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

    International Nuclear Information System (INIS)

    Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

    2002-01-01

    Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

  7. JSD1000: multi-group cross section sets for shielding materials

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)

  8. A contribution to shielding effectiveness analysis of shielded tents

    Directory of Open Access Journals (Sweden)

    Vranić Zoran M.

    2004-01-01

    Full Text Available An analysis of shielding effectiveness (SE of the shielded tents made of the metallised fabrics is given. First, two electromagnetic characteristic fundamental for coupling through electrically thin shield, the skin depth break frequency and the surface resistance or transfer impedance, is defined and analyzed. Then, the transfer function and the SE are analyzed regarding to the frequency range of interest to the Electromagnetic Compatibility (EMC Community.

  9. Effect of neutrons scattered from boundary of neutron field on shielding experiment

    International Nuclear Information System (INIS)

    Ogawa, Tatsuhiko; Abe, Takuya; Kosako, Toshiso; Iimoto, Takeshi

    2009-01-01

    Neutron shielding experiment with 49 cm-thick ordinary concrete was carried out at the reactor 'Yayoi' The University of Tokyo. System of this experiment is enclosed by heavy concrete where neutrons backscattered from heavy concrete likely affected neutron flux on the back surface of shielding concrete. Reaction rate of 197 Au(n, γ), cadmium covered 197 Au(n, γ) and 115 In(n, n') in the shielding concrete was measured using foil activation method. Neutron transport calculation was carried out in order to simulate reaction rate by calculating neutron spectra and convoluting with neutron capture cross-section in neutron shielding concrete. Comparison was made between calculated reaction rate and experimental one, and almost satisfactory agreement was found except for the back surface of shielding. To compose adequate simulation model, description of heavy concrete behind the shielding was thought to be of importance. For example, disregarding neutrons backscattered from heavy concrete, calculation underestimated reaction rate by the factor of 10. In another example, assuming that chemical composition of heavy concrete is equal to the composition adopted from a literature, the reaction rate was overestimated by factor of 5. By making the composition of heavy concrete equal to that based on facility design, overestimation was found to be the factor of 2. Therefore, adequate description of chemical composition of heavy concrete is found to be of importance in order to simulate neutron induced reaction rate on the back surface of neutron shielding concrete in shielding experiment performed in a system enclosed by heavy concrete. (author)

  10. DECOMMISSIONING OF SHIELDED FACILITIES AT WINFRITH USED FOR POST IRRADIATION EXAMINATION OF NUCLEAR FUELS and OTHER ACTIVE ITEMS

    International Nuclear Information System (INIS)

    Miller, K.D.; Parkinson, S.J.; Cornell, R.M.; Staples, A.T.

    2003-01-01

    This paper describes the approaches used in the clearing, cleaning, decontamination and decommissioning of a very large suite of seven concrete shielded caves and other facilities used by UKAEA at Winfrith Technology Centre, England over a period of about 30 years for the postirradiation examination (PIE) of a wide range of nuclear fuels and other very active components. The basic construction of the facilities will first be described, setting the scene for the major challenges that 1970s' thinking posed for decommissioning engineers. The tendency then to use large and heavy items of equipment supported upon massive steel bench structures produced a series of major problems that had to be overcome. The means of solving these problems by utilization of relatively simple and inexpensive equipment will be described. Later, a further set of challenges was experienced to decontaminate the interior surfaces to allow man entries to be undertaken at acceptable dose rates. The paper will describe the types of tooling used and the range of complementary techniques that were employed to steadily reduce the dose rates down to acceptable levels. Some explanations will also be given for the creation of realistic dose budgets and the methods of recording and continuously assessing the progress against these budgets throughout the project. Some final considerations are given to the commercial approaches to be adopted throughout this major project by the decommissioning engineers. Particular emphasis will be given to the selection of equipment and techniques that are effective so that the whole process can be carried out in a cost-effective and timely manner. The paper also provides brief complementary information obtained during the decommissioning of a plutonium-contaminated facility used for a range of semi-experimental purposes in the late 1970s. The main objective here was to remove the alpha contamination in such a manner that the volume of Plutonium Contaminated Materials (P

  11. Transparent fast neutron shielding material and shielding method

    International Nuclear Information System (INIS)

    Nashimoto, Tetsuji; Katase, Haruhisa.

    1993-01-01

    Polyisobutylene having a viscosity average molecular weight of 20,000 to 80,000 and a hydrogen atom density of greater than 7.0 x 10 22 /cm 3 is used as a fast neutron shielding material. The shielding material is excellent in the shielding performance against fast neutrons, and there is no worry of leakage even when holes should be formed to a vessel. Further, it is excellent in fabricability, relatively safe even upon occurrence of fire and, in addition, it is transparent to enable to observe contents easily. (T.M.)

  12. Electromagnetically shielded building

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1992-04-21

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

  13. Electromagnetically shielded building

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1992-01-01

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  14. 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.)

  15. Shielding augmentation of roll-on shield from NAPS to Kaiga-2

    International Nuclear Information System (INIS)

    Pradhan, A.S.; Kumar, A.N.

    2000-01-01

    Extensive radiation field surveys were conducted in NAPS and KAPS reactor buildings as a part of commissioning checks on radiation shielding. During such surveys, dose rate higher than the expected values were noticed in fuelling machine service areas. A movable shield, separating high field area fuelling machine vault and low field area fuelling machine service area, known as roll-on shield was identified as one of the causes of high field in fuelling machine service area along with weaker end-shield. This paper discusses systematic approach adopted in bringing down the dose rates in fuelling machine service area by augmentation of roll-on shield. (author)

  16. Gamma ray shielding: a web based interactive program

    International Nuclear Information System (INIS)

    Subbaiah, K.V.; Senthi Kumar, C.; Sarangapani, R.

    2005-01-01

    A web based interactive computing program is developed using java for quick assessment of Gamma Ray shielding problems. The program addresses usually encountered source geometries like POINT, LINE, CYLINDRICAL, ANNULAR, SPHERICAL, BOX, followed by 'SLAB' shield configurations. The calculation is based on point kernel technique. The source points are randomly sampled within the source volume. From each source point, optical path traversed in the source and shield media up to the detector location is estimated to calculate geometrical and material attenuations, and then corresponding buildup factor is obtained, which accounts for scattered contribution. Finally, the dose rate for entire source is obtained by summing over all sampled points. The application allows the user to select one of the seven regular geometrical bodies and provision exist to give source details such as emission energies, intensities, physical dimensions and material composition. Similar provision is provided to specify shield slab details. To aid the user, atomic numbers, densities, standard build factor materials and isotope list with respective emission energies and intensity for ready reference are given in dropdown combo boxes. Typical results obtained from this program are validated against existing point kernel gamma ray shielding codes. Additional facility is provided to compute fission product gamma ray source strengths based on the fuel type, burn up and cooling time. Plots of Fission product gamma ray source strengths, Gamma ray cross-sections and buildup factors can be optionally obtained, which enable the user to draw inference on the computed results. It is expected that this tool will be handy to all health physicists and radiological safety officers as it will be available on the internet. (author)

  17. Wake Shield Target Protection

    International Nuclear Information System (INIS)

    Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

    2003-01-01

    The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

  18. 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)

  19. 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)

  20. Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

    International Nuclear Information System (INIS)

    Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

    2007-01-01

    X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement. Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities

  1. Survey of shielding calculation parameters in radiotherapy rooms used in the country and its impact in the existing calculation methodologies

    International Nuclear Information System (INIS)

    Japiassu, Fernando Parois

    2013-01-01

    When designing radiotherapy treatment rooms, the dimensions of barriers are established on the basis of American calculation methodologies specifically; NCRP Report N° 49, NCRP Report N° 51, and more recently, NCRP Report N° 151. Such barrier calculations are based on parameters reflecting predictions of treatments to be performed within the room; which, in tum, reftect a specific reality found in a country. There exists, however, a variety of modern radiotherapy techniques, such as Intensity Modulated Radiation Therapy (IMRT); Total Body Irradiation (TBl) and radiosurgery (SRS); where patierits are treated in a much different way than during more conventional treatrnents, which are not taken into account the traditional shielding calculation methodology. This may lead to a faulty design of treattnent rooms. In order to establish a comparison between the methodology used to calculate shielding design and the reality of treatments performed in Brazil, two radiotherapy facilitie were selected, both of them offering traditional and modern treatment techniqued as described above. Data in relation with reatments perfotmed over a period of six (6)months of operations in both institutions were collected. Based on tlis informaton, a new set of realistic parameters required for shielding design was estãblished, whicb in turn allowed for a nwe caculation of barrier thickness for both facilities. The barrier thickness resultaing from this calculation was then compared with the barrier thickness propose as part of the original shielding design, approved by the regulatory authority. First, concerning the public facility, the thickness of all primary barriers proposed in the shielding design was actually larger than the thickness resulting from calculations based on realistic parameters. Second, concerning the private facility, the new data show that the thickness of three out of the four primary barriers described in the project is larger than the thickness oresulting from

  2. Optimal beta-ray shielding thicknesses for different therapeutic radionuclides and shielding materials

    International Nuclear Information System (INIS)

    Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

    2017-01-01

    To better understand the distribution of deposited energy of beta and gamma rays according to changes in shielding materials and thicknesses when radionuclides are used for therapeutic nuclear medicine, a simulation was conducted. The results showed that due to the physical characteristics of each therapeutic radionuclide, the thicknesses of shielding materials at which beta-ray shielding takes place varied. Additional analysis of the shielding of gamma ray was conducted for radionuclides that emit both beta and gamma rays simultaneously with results showing shielding effects proportional to the atomic number and density of the shielding materials. Also, analysis of bremsstrahlung emission after beta-ray interactions in the simulation revealed that the occurrence of bremsstrahlung was relatively lower than theoretically calculated and varied depending on different radionuclides. (authors)

  3. Shielding calculations for the TFTR neutral beam injectors

    International Nuclear Information System (INIS)

    Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1979-07-01

    Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

  4. Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki

    2007-01-01

    The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)

  5. Shielding synchrotron light sources: Advantages of circular shield walls tunnels

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.L. [Design and Accelerator Operations Consulting, 568 Wintergreen Ct Ridge, NY 11961 (United States); Ghosh, V.J.; Breitfeller, M. [NSLS-II, Brookhaven National Laboratory, Upton, NY 11973 (United States)

    2016-08-11

    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 produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

  6. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Katsuhiko; Maekawa, Fujio; Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated. (author)

  7. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    CERN Document Server

    Yoshida, K; Takada, H

    2003-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated.

  8. An evaluation of NCRP report 151--radiation shielding design for radiotherapy facilities, and a feasibility study for 6 MV open-door treatments in an existing high-energy radiation therapy bunker

    Science.gov (United States)

    Kildea, John

    This thesis describes a study of shielding design techniques used for radiation therapy facilities that employ megavoltage linear accelerators. Specifically, an evaluation of the shielding design formalism described in NCRP report 151 was undertaken and a feasibility study for open-door 6 MV radiation therapy treatments in existing 6 MV, 18 MV treatment rooms at the Montreal General Hospital (MGH) was conducted. To evaluate the shielding design formalism of NCRP 151, barrier-attenuated equivalent doses were measured for several of the treatment rooms at the MGH and compared with expectations from NCRP 151 calculations. It was found that, while the insight and recommendations of NCRP 151 are very valuable, its dose predictions are not always correct. As such, the NCRP 151 methodology is best used in conjunction with physical measurements. The feasibility study for 6 MV open-door treatments made use of the NCRP 151 formalism, together with physical measurements for realistic 6 MV workloads. The results suggest that, dosimetrically, 6 MV open door treatments are feasible. A conservative estimate for the increased dose at the door arising from such treatments is 0.1 mSv, with a 1/8 occupancy factor, as recommended in NCRP 151, included.

  9. Shield requirement estimation for pin storage room in fuel fabrication plant

    International Nuclear Information System (INIS)

    Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

    2012-01-01

    Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

  10. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1989-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating (ICRH) operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility (RFTF), a magnetic mirror device at Oak Ridge National Laboratory (ORNL), using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages (∼500 A, ∼20 kV at 25 kW) within 50% of those expected in tokamaks. The time varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with a Langmuir probe. Both probes were scanned in front of the antenna. Ion energies were measured with a gridded energy analyzer located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurements show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Plasma parameters and ion energies have been correlated with the antenna current and used in s computational model of the plasma sheath to predict the amount of erosion expected from the Faraday shield elements exposed to plasma. Predictions of light ion sputtering of candidate Faraday shield materials are presented. 19 refs., 6 figs., 1 tab

  11. Decontamination of an Analytical Laboratory Hot Cell Facility

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; Rosenberg, K.E.; Coleman, R.M.

    1995-11-01

    An Analytical Laboratory Hot Cell Facility at Argonne National Laboratory-West (ANL-W) had been in service for nearly thirty years. In order to comply with current DOE regulations governing such facilities and meet programmatic requirements, a major refurbishment effort was mandated. Due to the high levels of radiation and contamination within the cells, a decontamination effort was necessary to provide an environment that permitted workers to enter the cells to perform refurbishment activities without receiving high doses of radiation and to minimize the potential for the spread of contamination. State-of-the-art decontamination methods, as well as time-proven methods were utilized to minimize personnel exposure as well as maximize results

  12. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

    The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

  13. Analysis of the JASPER Program Radial Shield Attenuation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1993-01-01

    The results of the analysis of the JASPER Program Radial Shield Attenuation Experiment are presented. The experiment was performed in 1986 at the ORNL Tower Shielding Facility. It is the first of six experiments in this cooperative Japanese and American program in support of shielding designs for advanced sodium-cooled reactors. Six different shielding configurations and subconfigurations thereof were studied. The configurations were calculated with the DOT-IV two-dimensional discrete ordinates radiation transport computer code using the R-Z geometry option, a symmetric S{sub 12} quadrature (96 directions), and cross sections from ENDF/B versions IV and V in either a 51- or 61-group structure. Auxiliary codes were used to compute detector responses and prepare cross sections and source input for the DOT-IV calculations. Calculated detector responses were compared with measured responses and the agreement was good to excellent in many cases. However, the agreement for configurations having thick steel or B{sub 4}C regions or for some very large configurations was fair to poor. The disagreement was attributed to cross-section data, broad-group structure, or high background in the measurements. In particular, it is shown that two cross-section sets for ``B give very different results for neutron transmission through the thick B{sub 4}C regions used in one set of experimental configurations. Implications for design calculations are given.

  14. Makers of clotting medicine not entitled to liability shield.

    Science.gov (United States)

    1996-02-23

    The Indiana Court of Appeals, Second District, reversed a ruling that protected the manufacturer of a hemophilia blood clotting medicine from legal liability. A boy with hemophilia, known in court records as J.K.B. Jr., contracted HIV from the blood clotting medicine, Factor VIII. His parents sued four pharmaceutical companies, Armour Pharmaceutical Company, Cutter Laboratories, Baxter Healthcare Corporation, and Alpha Therapeutics Corporation, alleging that their products caused J.K.B. Jr.'s death. The companies convinced a Superior Court judge that they were protected under the Indiana Blood Shield Statute. The appeals court ruled that pharmaceutical companies are not included in the Blood Shield Statute. The statute was originally established to protect state-approved blood banks, hospitals, and blood storage facilities because they provide a vital public service.

  15. Neutron shieldings

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1979-01-01

    Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

  16. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  17. Clean room installations in a radiopharmaceutical facility

    International Nuclear Information System (INIS)

    2000-01-01

    The standards of radiopharmaceuticals on the facility, working environment and preparation control strategy are yet to be generated. In general, radiopharmaceuticals have short half-lives and emit gamma radiation. Due to its unique characteristics, its preparation has to be made in the fume hood and hot cell to avoid radiation exposure to workers. Considering radiation protection, the working environment has to be maintained under negative pressure so that dispersion of radiopharmaceuticals should be avoided. On the contrary, a positively pressurized working environment gives clean atmosphere and prevents contamination with harmful microorganisms during preparation. Hence, it is required to harmonize for mentioned contradictory conditions in preparation of radiopharmaceuticals for the safety of workers and its quality assurance as well. Therefore, it is reasonable that good manufacturing practice for radiopharmaceutical production facility should be constituted according to the standards for production of biological agents accompanied with a radiation shielding

  18. Pre-installation empirical testing of room shielding for high dose rate remote afterloaders

    International Nuclear Information System (INIS)

    Klein, E.E.; Grigsby, P.W.; Williamson, J.F.; Meigooni, A.S.

    1993-01-01

    PURPOSE: Many facilities are acquiring high dose rate remote afterloading units. It is economical that these units be placed in existing shielded teletherapy rooms. Scatter-radiation barriers marginally protect uncontrolled areas from a high dose rate source especially in a room that houses a non-dynamic Cobalt-60 unit. In addition the exact thickness and material composition of the barriers are unknown and therefore, a calculation technique may give misleading results. Also, it would be impossible to evaluate an entire wall barrier by taking isolated core samples in order to assist in the calculations. A quick and inexpensive measurement of dose equivalent using a rented high activity 192Ir source evaluates the barriers and locates shielding deficiencies. METHODS AND MATERIALS: We performed transmission calculations for primary and scattered radiation based on National Council on Radiation Protection and Measurements Reports 49 and 51, respectively. We then rented a high activity 21.7 Ci (8.03 x 10(11) Bq) Ir-192 source to assess our existing teletherapy room shielding for adequacy and voids. This source was placed at the proposed location for clinical high dose rate treatment and measurements were performed. RESULTS: No deficiencies were found in controlled areas surrounding the room, but large differences were found between the calculated and measured values. Our survey located a region in the uncontrolled area above the room requiring augmented shielding which was not predicted by the calculations. A canopy shield was designed to potentially augment the shielding in the ceiling direction. CONCLUSION: Pre-installation testing by measurement is an invaluable method for locating shielding deficiencies and avoiding unnecessary enhancement of shielding particularly when there is lack of information of the inherent shielding

  19. Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

    International Nuclear Information System (INIS)

    Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

    1981-01-01

    The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

  20. Performance, Facility Pressure Effects, and Stability Characterization Tests of NASA's Hall Effect Rocket with Magnetic Shielding Thruster

    Science.gov (United States)

    Kamhawi, Hani; Huang, Wensheng; Haag, Thomas; Yim, John; Herman, Daniel; Williams, George; Gilland, James; Peterson, Peter; Hofer, Richard; Mikellides, Ioannis

    2016-01-01

    NASAs Hall Effect Rocket with Magnetic Shielding (HERMeS) 12.5 kW Technology Demonstration Unit-1 (TDU-1) Hall thruster has been the subject of extensive technology maturation in preparation for flight system development. Part of the technology maturation effort included experimental evaluation of the TDU-1 thruster with conducting and dielectric front pole cover materials in two different electrical configurations. A graphite front pole cover thruster configuration with the thruster body electrically tied to cathode and an alumina front pole cover thruster configuration with the thruster body floating were evaluated. Both configurations were also evaluated at different facility background pressure conditions to evaluate background pressure effects on thruster operation. Performance characterization tests found that higher thruster performance was attained with the graphite front pole cover configuration with the thruster electrically tied to cathode. A total thrust efficiency of 68 and a total specific impulse of 2,820 s was demonstrated at a discharge voltage of 600 V and a discharge power of 12.5 kW. Thruster stability regimes were characterized with respect to the thruster discharge current oscillations and with maps of the current-voltage-magnetic field (IVB). Analysis of TDU-1 discharge current waveforms found that lower normalized discharge current peak-to-peak and root mean square magnitudes were attained when the thruster was electrically floated with alumina front pole covers. Background pressure effects characterization tests indicated that the thruster performance and stability was mostly invariant to changes in the facility background pressure for vacuum chamber pressure below 110-5 Torr-Xe (for thruster flow rate above 8 mgs). Power spectral density analysis of the discharge current waveform showed that increasing the vacuum chamber background pressure resulted in a higher discharge current dominant frequency. Finally the IVB maps of the TDU-1

  1. Facilities for Research and Development of Medical Radioisotopes

    International Nuclear Information System (INIS)

    Shin, Byung Chul; Choung, Won Myung; Park, Jin Ho

    2003-03-01

    This study is carried out by KAERI(Korea Atomic Energy Research Institute) to construct the basic facilities for development and production of medical radioisotope. For the characteristics of radiopharmaceuticals, the facilities should be complied with the radiation shield and GMP(Good Manufacturing Practice) guideline. The KAERI, which has carried out the research and development of the radiopharmaceuticals, made a design of these facilities and built them in the HANARO Center and opened the technique and facilities to the public to give a foundation for research and development of the radiopharmaceuticals. In the facilities, radiation shielding utilities and GMP instruments were set up and their operating manuals were documented. Every utilities and instruments were performed the test to confirm their efficiency and the approval for use of the facilities will be achieved from MOST(Ministry of Science and Technology). It is expected to be applied in development of therapeutic radioisotope such as Re-188 generator and Ho-166, as well as Tc-99m generator and Sr-89 chloride for medical use. And it also looks forward to the contribution to the related industry through the development of product in high demand and value

  2. 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.)

  3. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

  4. The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the SILOE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  5. The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the OSIRIS reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  6. Electromagnetic shield

    International Nuclear Information System (INIS)

    Miller, J.S.

    1987-01-01

    An electromagnetic shield is described comprising: closed, electrically-conductive rings, each having an open center; and binder means for arranging the rings in a predetermined, fixed relationship relative to each other, the so-arranged rings and binder means defining an outer surface; wherein electromagnetic energy received by the shield from a source adjacent its outer surface induces an electrical current to flow in a predetermined direction adjacent and parallel to the outer surface, through the rings; and wherein each ring is configured to cause source-induced alternating current flowing through the portion of the ring closest to the outer surface to electromagnetically induce an oppositely-directed current in the portion of the ring furthest from the surface, such oppositely-directed current bucking any source-induced current in the latter ring portion and thus reducing the magnitude of current flowing through it, whereby the electromagnetic shielding effected by the shield is enhanced

  7. A shielding chamber for the Rossendorf whole body counter; Eine Abschirmkammer fuer den Rossendorfer Ganzkoerperzaehler

    Energy Technology Data Exchange (ETDEWEB)

    Beutmann, A.; Ebert, S.; Kaden, M.; Loehnert, D. [VKTA, Dresden (Germany); Doerfel, H.R. [IDEA System GmbH, Karlsruhe (Germany); Schreiber, W. [Ingenieurbedarf G. Schoene und W. Schreiber GmbH, Grosserkmannsdorf (Germany); Helbig, S.

    2016-07-01

    In connection with the relocation of the incorporation measurement point operated by the VKTA, a new shielding chamber was designed. The development of the new shielding chamber will be shown based on the design study by IDEA System and the inquiries for material availability, procurement of material and assembly technology up to fabrication, assembly and completion of the chamber. The accompanying background measurements through In-situ gamma spectrometry and first experiences with incorporation measurements at the new In-Vivo measurement facility are shown.

  8. Effect of CSR shielding in the compact linear collider

    CERN Document Server

    Esberg, J; Apsimon, R; Schulte, D

    2014-01-01

    The Drive Beam complex of the Compact Linear Collider must use short bunches with a large charge making beam transport susceptible to unwanted effects of Coherent Synchrotron Radiation emitted in the dipole magnets. We present the effects of transporting the beam within a limited aperture which decreases the magnitude of the CSR wake. The effect, known as CSR shielding, eases the design of key components of the facility.

  9. Intrinsic noise of a superheated droplet detector for neutron background measurements in massively shielded facilities

    Directory of Open Access Journals (Sweden)

    Fernandes Ana C.

    2017-01-01

    Full Text Available Superheated droplet detectors are a promising technique to the measurement of low-intensity neutron fields, as detectors can be rendered insensitive to minimum ionizing radiations. We report on the intrinsic neutron-induced signal of C2ClF5 devices fabricated by our group that originate from neutron- and alpha-emitting impurities in the detector constituents. The neutron background was calculated via Monte Carlo simulations using the MCNPX-PoliMi code in order to extract the recoil distributions following neutron interaction with the atoms of the superheated liquid. Various nuclear techniques were employed to characterise the detector materials with respect to source isotopes (238U, 232Th and 147Sm for the normalisation of the simulations and also light elements (B, Li having high (α, n neutron production yields. We derived a background signal of ~10-3 cts/day in a 1 liter detector of 1-3 wt.% C2ClF5, corresponding to a detection limit in the order of 10-8 n cm-2s-1. Direct measurements in a massively shielded underground facility for dark matter search have confirmed this result. With the borosilicate detector containers found to be the dominant background source in current detectors, possibilities for further noise reduction by ~2 orders of magnitude based on selected container materials are discussed.

  10. Shielding Calculations for PUSPATI TRIGA Reactor (RTP) Fuel Transfer Cask with Micro shield

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

    2011-01-01

    The shielding calculations for RTP fuel transfer cask was performed by using computer code Micro shield 7.02. Micro shield is a computer code designed to provide a model to be used for shielding calculations. The results of the calculations can be obtained fast but the code is not suitable for complex geometries with a shielding composed of more than one material. Nevertheless, the program is sufficient for As Low As Reasonable Achievable (ALARA) optimization calculations. In this calculation, a geometry based on the conceptual design of RTP fuel transfer cask was modeled. Shielding material used in the calculations were lead (Pb) and stainless steel 304 (SS304). The results obtained from these calculations are discussed in this paper. (author)

  11. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  12. Evaluation of Heat Shields from RTS Wright Industries Magnesium and Uranium Beds

    CERN Document Server

    Korinko, P S

    2002-01-01

    Heat shields from a factory test of the furnaces that will be used to heat the magnesium and uranium beds for the tritium extraction facility (TEF) were examined to determine the cause of discoloration. The samples were examined using visual, optical microscopy, electron microscopy, x-ray spectroscopy, and Auger electron spectroscopy.

  13. ALARA implementation in 131I therapeutic capsule production facility

    International Nuclear Information System (INIS)

    Kumawat, Lalit; Swaminathan, N.; Sudheer, T.S.; Sachdev, S.S.; Arora, S.S.; Vairalkar, K.G.

    2005-01-01

    Sodium iodide 131 I solution had been invariably administered to patients for both diagnosis and therapy of thyrotoxicosis. The undue exposure to non-target organs has been over come by introducing NaI ( 131 I) in a gelatin capsule. BRIT has set up experimental facility for the preparation and the production volume has augmented into four fold due to increase in demand and the same facility is being used to cater the need. However, the adequately shielded facility (fume hood) used for (manual) dispensing activity in capsules, capsules and product vial capping, transfer of the vials into lead pots and activity measurement of each vial has resulted in significant increase in the personnel exposure. The sources had been identified and efforts were made to reduce the exposure in these operations. An annular shield was introduced around the dispenser, resulted in the reduction of radiation field at wrist level by a factor of three. Introduction of shielded automated dispenser and usage of longer tools for transfer and capping of vials has effected in two times reduction of collective wrist dose. Currently, the relocated capping station two meters away from the source certainly will bring down further exposure. (author)

  14. 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.

  15. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1989-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility, a magnetic mirror device at Oak Ridge National Laboratory, using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages within 50% of those expected in tokamaks. The time-varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with Langmuir probe. Both probes were scanned in front of the antenna. Ion energies were measured with a gridded energy analyzer located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurement show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Electron temperatures increase with rf power and can reach values ≥60 eV for an rf power of ∼25 kW. Incident ion energies ≥300 eV have been measured for the same power level. Predictions of light ion sputtering of candidate Faraday shield materials are presented. 19 refs., 6 figs., 1 tab

  16. Design of emergency shield

    International Nuclear Information System (INIS)

    Soliman, S.E.

    1993-01-01

    Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

  17. RadShield: semiautomated shielding design using a floor plan driven graphical user interface.

    Science.gov (United States)

    DeLorenzo, Matthew C; Wu, Dee H; Yang, Kai; Rutel, Isaac B

    2016-09-08

    The purpose of this study was to introduce and describe the development of RadShield, a Java-based graphical user interface (GUI), which provides a base design that uniquely performs thorough, spatially distributed calculations at many points and reports the maximum air-kerma rate and barrier thickness for each barrier pursuant to NCRP Report 147 methodology. Semiautomated shielding design calculations are validated by two approaches: a geometry-based approach and a manual approach. A series of geometry-based equations were derived giv-ing the maximum air-kerma rate magnitude and location through a first derivative root finding approach. The second approach consisted of comparing RadShield results with those found by manual shielding design by an American Board of Radiology (ABR)-certified medical physicist for two clinical room situations: two adjacent catheterization labs, and a radiographic and fluoroscopic (R&F) exam room. RadShield's efficacy in finding the maximum air-kerma rate was compared against the geometry-based approach and the overall shielding recommendations by RadShield were compared against the medical physicist's shielding results. Percentage errors between the geometry-based approach and RadShield's approach in finding the magnitude and location of the maximum air-kerma rate was within 0.00124% and 14 mm. RadShield's barrier thickness calculations were found to be within 0.156 mm lead (Pb) and 0.150 mm lead (Pb) for the adjacent catheteriza-tion labs and R&F room examples, respectively. However, within the R&F room example, differences in locating the most sensitive calculation point on the floor plan for one of the barriers was not considered in the medical physicist's calculation and was revealed by the RadShield calculations. RadShield is shown to accurately find the maximum values of air-kerma rate and barrier thickness using NCRP Report 147 methodology. Visual inspection alone of the 2D X-ray exam distribution by a medical physicist may not

  18. Radiation field characterization and shielding studies for the ELI Beamlines facility

    Czech Academy of Sciences Publication Activity Database

    Ferrari, A.; Amato, E.; Margarone, Daniele; Cowan, T.; Korn, Georg

    2013-01-01

    Roč. 272, May (2013), s. 138-144 ISSN 0169-4332 R&D Projects: GA MŠk ED1.1.00/02.0061; GA MŠk EE.2.3.20.0087; GA ČR(CZ) GAP205/11/1165 Grant - others:ELI Beamlines(XE) CZ.1.05/1.1.00/02.0061; OP VK 2 LaserGen(XE) CZ.1.07/2.3.00/20.0087; AVČR(CZ) M100101210 Institutional support: RVO:68378271 Keywords : particle acceleration from laser-matter interaction * shielding * Monte Carlo * radiation protection Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.538, year: 2013

  19. Evaluation of the radiation field and shielding assessment of the experimental area of HIE-ISOLDE

    International Nuclear Information System (INIS)

    Romanets, Y.; Goncalves, I.F.; Maria, S. di; Vaz, P.; Vollaire, J.; Bernardes, A.P.; Dorsival, A.; Kadi, Y.; Vlachoudis, V.

    2014-01-01

    The ISOLDE facility at CERN is one of the first facilities in the world dedicated to the production of the radioactive ion beams (RIB) and during all its working time underwent several upgrades. The goal of the latest proposed upgrade, 'The High Intensity and Energy ISOLDE' (HIE-ISOLDE), is to provide a higher performance facility in order to approximate it to the level of the next generation ISOL facilities, like EURISOL. The HIE-ISOLDE aims to improve significantly the quality of the produced RIB and for this reason the increasing of the primary beam power is one of the main objectives of the project. An increase in the nominal beam current (from 2 to 6 μA proton beam intensity) and energy (from 1.4 GeV to 2 GeV) of the primary proton beam will be possible due to the upgrade of CERN's accelerator infrastructure. The current upgrade means reassessment of the radiation protection and the radiation safety of the facility. However, an evaluation of the existing shielding configuration and access restrictions to the experimental and supply areas must be carried out. Monte Carlo calculations were performed in order to evaluate the radiation protection of the facility as well as radiation shielding assessment and design. The FLUKA-Monte Carlo code was used in this study to calculate the ambient dose rate distribution and particle fluxes in the most important areas, such as the experimental hall of the facility. The results indicate a significant increase in the ambient dose equivalent rate in some areas of the experimental hall when an upgrade configuration of the primary proton beam is considered. Special attention is required for the shielding of the target area once it is the main and very intensive radiation source, especially under the upgrade conditions. In this study, the access points to the beam extraction and beam maintenance areas, such as the mass separator rooms and the high voltage room, are identified as the most sensitive for the experimental hall from

  20. 324 Facility B-Cell quality process plan

    International Nuclear Information System (INIS)

    Carlson, J.L.

    1998-01-01

    This report documents the quality process plan for the restart of a hot cell in the B Plant, originally a bismuth phosphate processing facility, but later converted to a waste fractionation plant. B-Cell is currently being cleaned out and deactivated. TPA Milestone M-89-02 dictates that all mixed waste and equipment be removed from B-Cell by 5/31/1999. This report describes the major activities that remain for completion of the TPA milestone

  1. Methodology for worker neutron exposure evaluation in the PDCF facility design

    International Nuclear Information System (INIS)

    Scherpelz, R. I.; Traub, R. J.; Pryor, K. H.

    2004-01-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y -1 for the whole body and 100 mSv y -1 for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were

  2. Internet accessible hot cell with gamma spectroscopy at the Missouri S and T nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grant, Edwin [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States); Mueller, Gary, E-mail: gmueller@mst.edu [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States); Castano, Carlos; Usman, Shoaib; Kumar, Arvind [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States)

    2011-08-15

    Highlights: > A dual-chambered internet-accessible heavily shielded facility has been built. > The facility allows distance users to analyze neutron irradiated samples remotely. > The Missouri S and T system uses computer automation with user feedback. > The system can analyze multiple samples and assist several researchers concurrently. - Abstract: A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S and T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently available for irradiation and analysis of samples. The facility allows authorized distance users engaged in collaborative activities with Missouri S and T to remotely manipulate and analyze neutron irradiated samples. The system consists of two shielded compartments, one for multiple sample storage, and the other dedicated exclusively for radiation measurements and spectroscopy. The second chamber has multiple detector ports, with graded shielding, and has the capability to support gamma spectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connected though a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based system complements the MSTR's current bare pneumatic tube (BPT) and cadmium lined pneumatic tube (CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g, irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals who do not have on-site access to a nuclear reactor facility to remotely participate in research and educational activities.

  3. Internet accessible hot cell with gamma spectroscopy at the Missouri S and T nuclear reactor

    International Nuclear Information System (INIS)

    Grant, Edwin; Mueller, Gary; Castano, Carlos; Usman, Shoaib; Kumar, Arvind

    2011-01-01

    Highlights: → A dual-chambered internet-accessible heavily shielded facility has been built. → The facility allows distance users to analyze neutron irradiated samples remotely. → The Missouri S and T system uses computer automation with user feedback. → The system can analyze multiple samples and assist several researchers concurrently. - Abstract: A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S and T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently available for irradiation and analysis of samples. The facility allows authorized distance users engaged in collaborative activities with Missouri S and T to remotely manipulate and analyze neutron irradiated samples. The system consists of two shielded compartments, one for multiple sample storage, and the other dedicated exclusively for radiation measurements and spectroscopy. The second chamber has multiple detector ports, with graded shielding, and has the capability to support gamma spectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connected though a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based system complements the MSTR's current bare pneumatic tube (BPT) and cadmium lined pneumatic tube (CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g, irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals who do not have on-site access to a nuclear reactor facility to remotely participate in research and educational activities.

  4. Neutron shielding material

    International Nuclear Information System (INIS)

    Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

    1993-01-01

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  5. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  6. 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

  7. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  8. Primary shield displacement and bowing

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    The reactor primary shield is constructed of high density concrete and surrounds the reactor core. The inlet, outlet and side primary shields were constructed in-place using 2.54 cm (1 in) thick steel plates as the forms. The plates remained as an integral part of the shields. The elongation of the pressure tubes due to thermal expansion and pressurization is not moving through the inlet nozzle hardware as designed but is accommodated by outward displacement and bowing of the inlet and outlet shields. Excessive distortion of the shields may result in gas seal failures, intolerable helium gas leaks, increased argon-41 emissions, and shield cooling tube failures. The shield surveillance and testing results are presented

  9. Evaluation of the shielding integrity of end-shields in PHWR type NPPs

    International Nuclear Information System (INIS)

    Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

    1996-01-01

    In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

  10. Magnetic shielding for superconducting RF cavities

    Science.gov (United States)

    Masuzawa, M.; Terashima, A.; Tsuchiya, K.; Ueki, R.

    2017-03-01

    Magnetic shielding is a key technology for superconducting radio frequency (RF) cavities. There are basically two approaches for shielding: (1) surround the cavity of interest with high permeability material and divert magnetic flux around it (passive shielding); and (2) create a magnetic field using coils that cancels the ambient magnetic field in the area of interest (active shielding). The choice of approach depends on the magnitude of the ambient magnetic field, residual magnetic field tolerance, shape of the magnetic shield, usage, cost, etc. However, passive shielding is more commonly used for superconducting RF cavities. The issue with passive shielding is that as the volume to be shielded increases, the size of the shielding material increases, thereby leading to cost increase. A recent trend is to place a magnetic shield in a cryogenic environment inside a cryostat, very close to the cavities, reducing the size and volume of the magnetic shield. In this case, the shielding effectiveness at cryogenic temperatures becomes important. We measured the permeabilities of various shielding materials at both room temperature and cryogenic temperature (4 K) and studied shielding degradation at that cryogenic temperature.

  11. Aggregation control of quantum dots through ion-mediated hydrogen bonding shielding.

    Science.gov (United States)

    Liu, Jianbo; Yang, Xiaohai; Wang, Kemin; He, Xiaoxiao; Wang, Qing; Huang, Jin; Liu, Yan

    2012-06-26

    Nanoparticle stabilization against detrimental aggregation is a critical parameter that needs to be well controlled. Herein, we present a facile and rapid ion-mediated dispersing technique that leads to hydrophilic aggregate-free quantum dots (QDs). Because of the shielding of the hydrogen bonds between cysteamine-capped QDs, the presence of F(-) ions disassembled the aggregates of QDs and afforded their high colloidal stability. The F(-) ions also greatly eliminated the nonspecific adsorption of the QDs on glass slides and cells. Unlike the conventional colloidal stabilized method that requires the use of any organic ligand and/or polymer for the passivation of the nanoparticle surface, the proposed approach adopts the small size and large diffusion coefficient of inorganic ions as dispersant, which offers the disaggregation a fast reaction dynamics and negligible influence on their intrinsic surface functional properties. Therefore, the ion-mediated dispersing strategy showed great potential in chemosensing and biomedical applications.

  12. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

    1995-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

  13. Shielding evaluation of neutron generator hall by Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

    2017-04-01

    A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

  14. {sup 3}He detector analysis of some special shielding materials

    Energy Technology Data Exchange (ETDEWEB)

    Avdic, S; Pesic, M [Boris Kidric, Institute of Nuclear Sciences, Beograd (Yugoslavia); Marinkovic, P [ETF Belgrade Univ. (Yugoslavia)

    1990-07-01

    The shielding properties of commercial materials of reactor Experiments, Inc. (R/X) were analyzed at the facility which includes bare heavy water experimental reactor RB with external neutron converter ENC, The fast neutron spectrum measurements in energy range from 1 MeV to 10 MeV was performed using ORTEC semiconductor neutron detector with He{sup 3} in diode coincidence arrangement. The neutron spectra have been evaluated from measured pulse-height distribution using numerical code HE3 for computation of detector efficiency in a collimated neutron beam. The neutron dose rates behind ENC with and without sample R/X material were determined using cubic spline interpolation routine for calculating the corresponding flux-dose rate conversion factors. Satisfactory shielding properties of the examined material in a fast neutron field in measurements and calculations are demonstrated. (author)

  15. Shielding container for radioactive isotopes

    International Nuclear Information System (INIS)

    Sumi, Tetsuo; Tosa, Masayoshi; Hatogai, Tatsuaki.

    1975-01-01

    Object: To effect opening and closing bidirectional radiation used particularly for a gamma densimeter or the like by one operation. Structure: This device comprises a rotatable shielding body for receiving radioactive isotope in the central portion thereof and having at least two radiation openings through which radiation is taken out of the isotope, and a shielding container having openings corresponding to the first mentioned radiation openings, respectively. The radioactive isotope is secured to a rotational shaft of the shielding body, and the shielding body is rotated to register the openings of the shielding container with the openings of the shielding body or to shield the openings, thereby effecting radiation and cut off of gamma ray in the bidirection by one operation. (Kamimura, M.)

  16. Experimental Investigation of Multi-layer Insulation Effect on Damage of Stuffed Shield by High-velocity Impact

    Directory of Open Access Journals (Sweden)

    GUAN Gong-shun

    2016-09-01

    Full Text Available The stuffed shield with multi-layer insulation(MLI was designed by improving on Al Whipple shield, and a series of high-velocity impact tests were practiced with a two-stage light gas gun facility at vacuum environment. The damage model of the stuffed shield with different MLI location by Al-sphere projectile impacting was obtained. The effect of MLI on damage of the stuffed shield by high-velocity impact was studied. The results indicate when the MLI is located at front side of the first Al-plate, the protection performance of the stuffed shield is improved with the larger perforation diameter of the first Al-plate and more impact kinetic energy dissipation of the projectile. When MLI is arranged at back side of the first Al-plate, the expansion of the secondary debris cloud from projectile impacting the first Al-plate is restrained, it is not good to improve the protection performance of the stuffed shield. When MLI is arranged at front side of the stuffed wall, the perforation size of the stuffed wall increases; when MLI is arranged at front side of the rear wall, the distribution range of crater on the rear wall decreases.

  17. EVALUATION OF BRACHYTHERAPY FACILITY SHIELDING STATUS IN KOREA OBTAINED FROM RADIATION SAFETY REPORTS

    Directory of Open Access Journals (Sweden)

    MI HYUN KEUM

    2013-10-01

    Full Text Available Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4% included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 (R-m2/Ci·hr, as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.

  18. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  19. Shielding in experimental areas

    International Nuclear Information System (INIS)

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  20. 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

  1. 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.

  2. Equipment experience in a radioactive LFCM [liquid-fed ceramic melter] vitrification facility

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.; Dierks, R.D.; Sevigny, G.J.; Goles, R.W.; Surma, J.E.; Thomas, N.M.

    1986-11-01

    Since October 1984, the Pacific Northwest Laboratory (PNL) has operated a pilot-scale radioactive liquid-fed ceramic melter (RLFCM) vitrification process in shielded manipulator hot cells. This vitrification facility is being operated for the Department of Energy (DOE) to remotely test vitrification equipment components in a radioactive environment and to develop design and operation data that can be applied to production-scale projects. This paper summarizes equipment and process experience obtained from the operations of equipment systems for waste feeding, waste vitrification, canister filling, canister handling, and vitrification off-gas treatment

  3. Shielding effects of concrete and foam external pipeline coatings

    International Nuclear Information System (INIS)

    Barlo, T.J.; Werner, D.P.

    1992-01-01

    The research project began in July, 1986 and was completed in December, 1990. The objectives of the research were: To determine whether concrete and urethane foam-barrier coatings shield the pipe from cathodic-protection current, To determine whether the barrier coatings also effectively shield the pipe from the environment, thus reducing the need for cathodic protection, To determine what levels of cathodic protection will be required to overcome shielding, and To establish what types of barrier coatings are most compatible with obtaining adequate levels of cathodic protection. To achieve these objectives, laboratory experiments were conducted with five barrier coating materials. These materials were (1) 2-lb/ft 3 , closed-cell urethane foam, (2) 3-lb/ft 3 , closed-cell urethane foam, (3) concrete barrier material, (4) glass fiber-reinforced concrete barrier material, and (5) sand. The barrier materials, whole and intentionally cracked, were applied to the bare, FBE-coated, and tape-coated steel specimens. The specimens were tested in aqueous electrolytes at room temperature and 140 degree F with no protection, protection to -0.95 V, and overprotection to -1.2 V (Cu/CuSO 4 )

  4. The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DIDO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  5. 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

  6. Transparent Conducting Graphene Hybrid Films To Improve Electromagnetic Interference (EMI) Shielding Performance of Graphene.

    Science.gov (United States)

    Ma, Limin; Lu, Zhengang; Tan, Jiubin; Liu, Jian; Ding, Xuemei; Black, Nicola; Li, Tianyi; Gallop, John; Hao, Ling

    2017-10-04

    Conducting graphene-based hybrids have attracted considerable attention in recent years for their scientific and technological significance in many applications. In this work, conductive graphene hybrid films, consisting of a metallic network fully encapsulated between monolayer graphene and quartz-glass substrate, were fabricated and characterized for their electromagnetic interference shielding capabilities. Experimental results show that by integration with a metallic network the sheet resistance of graphene was significantly suppressed from 813.27 to 5.53 Ω/sq with an optical transmittance at 91%. Consequently, the microwave shielding effectiveness (SE) exceeded 23.60 dB at the K u -band and 13.48 dB at the K a -band. The maximum SE value was 28.91 dB at 12 GHz. Compared with the SE of pristine monolayer graphene (3.46 dB), the SE of graphene hybrid film was enhanced by 25.45 dB (99.7% energy attenuation). At 94% optical transmittance, the sheet resistance was 20.67 Ω/sq and the maximum SE value was 20.86 dB at 12 GHz. Our results show that hybrid graphene films incorporate both high conductivity and superior electromagnetic shielding comparable to existing ITO shielding modalities. The combination of high conductivity and shielding along with the materials' earth-abundant nature, and facile large-scale fabrication, make these graphene hybrid films highly attractive for transparent EMI shielding.

  7. Nucleonic analysis of a preliminary design for the ETF neutral-beam-injector duct shielding

    International Nuclear Information System (INIS)

    Urban, W.T.; Seed, T.J.; Dudziak, D.J.

    1980-01-01

    A nucleonic analysis of the Engineering Test Facility Neutral-Beam-Injector duct shielding has been made using a hybrid Monte Carlo/discrete-ordinates method. This method used Monte Carlo to determine internal and external boundary surface sources for a subsequent discrete-ordinates calculation of the neutron and gamma-ray transport through the shield. The analysis also included determination of the energy and angular distribution of neutrons and gamma rays entering the duct from the torus plasma chamber. Confidence in the hybrid method and the results obtained were provided through a comparison with three-dimensional Monte Carlo results

  8. Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

    International Nuclear Information System (INIS)

    Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

    2002-01-01

    The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

  9. Damage sources for the NIF Grating Debris Shield (GDS) and methods for their mitigation

    Science.gov (United States)

    Carr, C. W.; Bude, J.; Miller, P. E.; Parham, T.; Whitman, P.; Monticelli, M.; Raman, R.; Cross, D.; Welday, B.; Ravizza, F.; Suratwala, T.; Davis, J.; Fischer, M.; Hawley, R.; Lee, H.; Matthews, M.; Norton, M.; Nostrand, M.; Vanblarcom, D.; Sommer, S.

    2017-11-01

    The primary sources of damage on the National Ignition Facility (NIF) Grating Debris Shield (GDS) are attributed to two independent types of laser-induced particulates. The first comes from the eruptions of bulk damage in a disposable debris shield downstream of the GDS. The second particle source comes from stray light focusing on absorbing glass armor at higher than expected fluences. We show that the composition of the particles is secondary to the energetics of their delivery, such that particles from either source are essentially benign if they arrive at the GDS with low temperatures and velocities.

  10. 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.)

  11. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-01-01

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  12. Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

    2004-10-01

    The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

  13. Guidelines for beamline and front-end radiation shielding design at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Fernandez, P.; X-Ray Science Division

    2008-01-01

    Shielding for the APS will be such that the individual radiation worker dose will be as low as reasonably achievable (ALARA). The ALARA goals for the APS are to keep the total of the work-related radiation exposure (exposure coming from other than natural or medical sources) as far below 500 person-mrem per year, collective total effective dose equivalent, as reasonably achievable. For an individual APS radiation worker, the goal is to keep the maximum occupational total effective dose equivalent of any one employee as far below 200 mrem/yr as reasonably achievable. The ALARA goal for APS beamline scientists is to keep the total of the work-related radiation exposure (exposure coming from other than natural or medical sources) as far below 100 person-mrem per year, collective total effective dose equivalent, as reasonably achievable. For an individual APS beamline scientist, the goal is to keep the maximum occupational total effective dose equivalent of any one scientist as far below 50 mrem/yr as reasonably achievable. The dose is actively monitored by the radiation monitors on the storage ring wall in each sector and by the frequent area surveys performed by the health physics personnel. For cases in which surveys indicate elevated hourly dose rates that may impact worker exposure, additional local shielding is provided to reduce the radiation field to an acceptable level. Passive area monitors are used throughout the facility to integrate doses in various areas. The results are analyzed for trends of increased doses, and shielding in these areas is evaluated and improved, as appropriate. The APS policy for on-site nonradiation workers in the vicinity of the APS facilities requires that the average nonradiation worker dose be below 0.2 mSv/yr (20 mrem/yr). In addition, the dose at the site boundary from all pathways is required to be below 0.1 mSv/yr (10 mrem/yr). For future modifications of the facility, the doses shall be evaluated and additional shielding

  14. Neutron field characterization and dosimetry at the TRIUMF proton therapy facility

    International Nuclear Information System (INIS)

    Mukherjee, B.

    2002-01-01

    Full text: In 1972 the 500 MeV H' Cyclotron of the TRIUMF (Tri University Meson Factory) located in Vancouver, Canada became operational. Beside Meson Physics, high-energy protons of various energy and beam current levels from the TRIUMF Cyclotron are used for scientific research and biomedical applications. Recently, a 500 MeV proton beam from the cyclotron was used as the booster beam for the radioactive ion beam facility, ISAC (Isotope Separator Accelerator) and a second beam as primary irradiation source for the Proton Irradiation Facility (PIF). The major commercial applications of the PIF are the provision of high-energy proton beams for radiation hardness testing of electronic components used in space applications (NASA) and proton therapy of ocular tumors (British Columbia Proton Therapy Facility). The PIF vault was constructed within the main accelerator hall of the TRIUMF using stacks of large concrete blocks. An intense field of fast neutrons is produced during the interaction of high-energy proton beam with target materials, such as, beam stops, collimators and beam energy degraders. The leakage of such neutrons due to insufficient radiological shielding or through the shielding discontinuities may constitute a major share of the personnel radiation exposure of the radiation workers. The neutron energy distribution and dose equivalent near a lead beam stopper bombarded with 116 MeV and 65 MeV collimated proton beams at the Ocular Tumor irradiation facility were evaluated using a Bonner-Sphere Spectrometer and a REM counter respectively. The results were utilized to investigate efficacy of the existing radiological shielding of the PIF. This paper highlights experimental methods to analyze the high-energy accelerator produced neutron beam and basic guideline for the radiological shielding designs of irradiation vault of Proton Therapy facilities

  15. The CERN-EU radiation facility for dosimetry at flight altitude and in space

    CERN Document Server

    Ferrari, A; Silari, Marco

    2001-01-01

    A reference facility for the inter-comparison of active and passive detectors in complex high-energy neutron fields is available at CERN since 1993. A positively charged hadron beam (a mixture of protons and pions) with momentum of 120 GeV/c hits a copper target, 50 cm thick and 7 cm in diameter. The secondary particles produced in the interaction traverse a shield made of either 80 cm of concrete or 40 cm of iron. Behind the iron shield, the resulting neutron spectrum has a maximum at about 1 MeV, with an additional high-energy component. Behind the concrete shield, the neutron spectrum has a second pronounced maximum at about 70 MeV and resembles the high- energy component of the radiation field at commercial flight altitudes created by cosmic rays. Recent Monte Carlo calculations are presented, performed for different beam conditions and shielding configurations in view of a possible upgrade of the facility for measurements related to the space program. (20 refs).

  16. The 'MELUSINE' reactor at Grenoble, France, and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the MELUSINE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities and specialized irradiation devices (loops and capsules). The information is presented in the form of six information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities

  17. Design and analysis of magnetic shield for 650 MHz SCRF cavity

    International Nuclear Information System (INIS)

    Thakur, Vanshree; Jain, Vikas; Das, S.; Shinde, R.S.; Joshi, S.C.

    2015-01-01

    Five-cell, 650 MHz Superconducting RF (SCRF) cavity is being developed at RRCAT for the Injector Linac of proposed ISNS project. The SCRF cavity needs to be shielded effectively from earth magnetic field. The external magnetic field can cause magnetic field trapping that limits the performance of SCRF cavity. The allowable limit of earth magnetic field in the cavity surface is < 10 mG. The magnetic shielding analysis carried out for 650 MHz dressed SCRF cavity is presented in this paper. For axial magnetic field shielding analysis, 2-D code PANDIRA has been used. A 2-D axisymmetric geometry (cylinder of Cryoperm10 sheet with 460 mm diameter of various thickness and 1100 mm length) has been modelled and analyzed in the presence of 240 mG external axial magnetic field. The influence of partial opening of 120 mm diameter at both ends of the cylinder on magnetic field pattern inside the shielded region has been evaluated. The transverse magnetic shielding analysis in the presence of 500 mG transverse external field has been carried out using OPERA 3D code. The flux leakage through the major openings for cavity supports, ports on the shield is investigated and accordingly the openings are designed to minimize the leakage. Inference of material thickness on the magnetic shielding for reducing magnetic field below specified limit has been investigated. Details of design and analysis of magnetic shield for SCRF cavity will be discussed in this paper. (author)

  18. CREST : a computer program for the calculation of composition dependent self-shielded cross-sections

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1977-01-01

    A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)

  19. Synthesis of mullite (3Al2O32SiO2) from local kaolin for radiation shielding

    Science.gov (United States)

    Ripin, Azuhar; Mohamed, Faizal; Aman, Asyraf

    2018-04-01

    Raw kaolin from Kota Tinggi, Johor was used in this study to produce ceramic mullite (3Al2O22SiO2) for radiation shielding materials. In this work, an attempt was made to study the potential of local minerals to be used as a shielding barrier for diagnostic radiology radiation facilities in hospitals and medical centers throughout Malaysia. The conventional ceramic processing route was employed in the study using different pressing strength and sintering time. The obtained samples were characterized using X-ray diffractometer (XRD) for phase identification of each of the samples. The lead equivalent (LE) test was carried out using 15.05 mCi Cobalt-57 with gamma energy of 122 keV to compute the abilities of the mullite ceramic samples to attenuate the radiation. XRD patterns of prepared ceramics revealed the presence of orthorhombic mullite, hexagonal quartz and orthorhombic sillimanite structures. Furthermore, the radiation test displayed the ability of ceramics to shield of 70 % of gamma radiation at the distance of 60 cm from the radiation source. The highest lead equivalent thickness is 1.0 mm Pb and the lowest is about 0.06 mm Pb. From the result, it is shown that the ceramic has the potential to use as a shielding barrier in diagnostic radiology facilities due to the ability of reducing the radiation dose up to 70 % from its initial value.

  20. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  1. Optimization of Photovoltaic Performance Through the Integration of Electrodynamic Dust Shield Layers

    Science.gov (United States)

    Nason, Steven; Davis, Kris; Hickman, Nicoleta; McFall, Judith; Arens, Ellen; Calle, Carlos

    2009-01-01

    The viability of photovoltaics on the Lunar and Martian surfaces may be determined by their ability to withstand significant degradation in the Lunar and Martian environments. One of the greatest threats is posed by fine dust particles which are continually blown about the surfaces. In an effort to determine the extent of the threat, and to investigate some abatement strategies, a series of experiments were conducted outdoors and in the Moon and Mars environmental chamber at the Florida Solar Energy Center. Electrodynamic dust shield prototypes based on the electric curtain concept have been developed by our collaborators at the Kennedy Space Center [1]. These thin film layers can remove dust from surfaces and prevent dust accumulation. Several types of dust shields were designed, built and tested under high vacuum conditions and simulated lunar gravity to validate the technology for lunar exploration applications. Gallium arsenide, single crystal and polycrystalline silicon photovoltaic integrated devices were designed, built and tested under Moon and Mars environmental conditions as well as under ambient conditions. Photovoltaic efficiency measurements were performed on each individual cell with the following configurations; without an encapsulation layer, with a glass covering, and with various thin film dust shields. It was found that the PV efficiency of the hybrid systems was unaffected by these various thin film dust shields, proving that the optical transmission of light through the device is virtually uninhibited by these layers. The future goal of this project is to incorporate a photovoltaic cell as the power source for the electrodynamic dust shield system, and experimentally show the effective removal of dust obstructing any light incident on the cell, thus insuring power production is maximized over time.

  2. Operation of post-irradiation examination facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eun Ka; Park, Kwang Joon; Jeon, Yong Bum [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-02-01

    In 1995, the post-irradiation examination (PIE) of nuclear fuels was performed as follows. The relation between burnup and top nozzle spring force of fuel assembly was obtained by measuring the holddown spring force on the Kori-1 reactor fuel assemblies. The resonance ultrasonic test for inspection of defect and moisture in fuel rod was carried out on fuel rods of C15 and J14 assemblies, and the change of fuel rod condition by storing in pool has been analyzed on the intentionally defected fuel rods (ID-C and ID-L) as well as intact fuel rod (1-2) by NDT in ht cell. The oxide layer thickness on cladding surface of J44-L12 fuel rod was measured by NDT method and metallography to reveal the oxidation as a function of temperature in the fuel rod, and the burnup of J44 fuel assembly was measured by chemical analysis. HVAC system and pool water treatment system of the PIE facility were continuously operated for air filtration and water purification. The monitoring of radiation and pool water in PIE facility has been carried out to maintain the facility safety, and electric power supply system was checked and maintained to supply the electric power to the facility normally. The developed measurement techniques of oxide layer thickness on fuel rod cladding and holddown spring force of top nozzle in fuel assembly were applied to examine the nuclear fuels. Besides, a radiation shielding glove box was designed and a hot cell compressor for volume reduction of radioactive materials was fabricated. 19 tabs., 38 figs., 7 refs. (Author) .new.

  3. Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

    International Nuclear Information System (INIS)

    Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

  4. Experience with Dismantling of the Analytic Cell in the JRTF Decommissioning Program

    International Nuclear Information System (INIS)

    Annoh, Akio; Nemoto, Koichi; Tajiri, Hideo; Saito, Keiichiro; Miyajima, Kazutoshi; Myodo, Masato

    2003-01-01

    The analytic cell was mainly used for process control analysis of the reprocessing process and for the measurement of fuel burn up ratio in JAERI's Reprocessing Test Facility (JRTF). The analytic cell was a heavy shielded one and equipped with a conveyor. The cell was alpha and beta(gamma)contaminated. For dismantling of analytic cells, it is very important to establish a method to remove the heavy shield safely and reduce the exposure. At first, a green house was set up to prevent the spread out of contamination, and next, the analytic cell was dismantled. Depending on the contamination condition, the workers wore protective suits such as air ventilated-suits for prevention of internal exposure and vinyl chloride aprons, lead aprons in order to reduce external exposure. From the work carried out, various data such as needed manpower for the activities, the collective dose of workers by external exposure, the amount of radioactive wastes and the relation between the weight of the shield and its dismantling efficiency were obtained and input for the database. The method of dismantling and the experience with the dismantling of the analytic cell in the JRTF, carried out during 2001 and 2002, are described in this paper

  5. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

    The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

  6. Background components of Ge(Li) and GeHP-detectors in the passive shield

    International Nuclear Information System (INIS)

    Buraeva, E.A.; Davydov, M.G.; Zorina, L.V.; Stasov, V.V.

    2007-01-01

    The gamma-spectrometer Ge(Li)- and the extra pure Ge-detector background components in a specially designed passive shield were subjected to investigation in the land-based laboratory in 1996-2006. The measurement time period varied from 45 up to 240 hours. The detector background is caused by the radionuclides in the shield material, in the shield cells and in the detector materials. The prominence was given to the study of the revealed time dependence of 222 Rn daughter product background including '2 10 Pb 46.5 keV peak [ru

  7. Use of closed circuit TV systems at the IEM cell

    International Nuclear Information System (INIS)

    Ames, C.P.

    1983-09-01

    The Interim Examination and Maintenance (IEM) Cell is a vertical, argon inerted hot cell located within the containment building of the Fast Flux Test Facility (FFTF). The IEM Cell has two basic functions: (1) to perform disassembly and nondestructive testing on radioactive, sodium-wetted reactor components, and (2) to provide a shielded work volume where radioactive, sodium-wetted reactor and refueling equipment can be repaired and maintained. Another important aspect of the cell is the ability to remotely remove, transfer and reinstall in-cell equipment for maintenance. This paper describes one technique used to enhance in-cell visibility during these remote operations

  8. 3-D analysis on arbitrarily-shaped ICRF antennas and Faraday shields

    International Nuclear Information System (INIS)

    Chen, G.L.; Whealton, J.H.; Baity, F.W.; Hoffman, D.J.; Owens, T.L.

    1986-01-01

    Cavity antennas with Faraday shields are proposed to couple ion cyclotron radio frequency power for heating fusion plasmas. This application requires small, high-power, low-frequency antennas. The results are presented of a theoretical study of the ICRF antennas being developed for this purpose at the Radio Frequency Test Facility (RFTF). The objectives of this work are to optimize experimental designs and to confirm test results

  9. Reconstruction of Co-60 Irradiation Facility No.1

    International Nuclear Information System (INIS)

    Nakamura, Yoshiteru; Takada, Isao; Kaneko, Hirohisa; Hirao, Toshio; Haneda, Noriyuki; Mitomo, Shouichi; Tachibana, Hiroyuki; Yoshida, Kenzou

    1989-01-01

    Cobalt Irradiation Facility No.1 was constructed in 1964 as the first large scale Co-60 irradiation facility equipped a deep water pool for source storage of Co-60 sources. Recently, the reconstruction of the facility was decided because the aging of various parts of the facility became remarkable and new research programs required upgradings of the facility. Important points of upgradings are as follows: A shielding capacity of the source storage and pool is increased to 55.5 PBq from 18.5 PBq. The opening in a floor of the irradiation room which is used for the source lifting in the room, is enlarged in order to utilize a large and high intensity source. Radiation resistance of the irradiation apparatus and installed equipments in the radiation room is increased for a high dose rate irradiation. Basic structure and shape of the facility building such as shielding, pool and building roof is not changed but electrical, mechanical equipments and systems are completely renewed. To increase a reliability, the irradiation apparatus and systems are also replaced with an improved and up-to-date one designed based on operation experiences of Co-60 facilities at TRCRE through many years. In addition, auxiliary equipments such as radiation monitors, manipulators, water treatment system and so on are replaced. This report presents the reconstruction of Co-60 Irradiation Facility No.1 stressing on the replacement and modification of the irradiation apparatus. (author)

  10. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions

  11. 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.

  12. 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)

  13. Development of silver coating process and facilities for ITER thermal shield

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D.K. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kim, R.G. [COTEC Corp., Changwon 641-846 (Korea, Republic of); Nam, K., E-mail: kwnam@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Noh, C.H.; Chung, W. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Yoon, D.C. [COTEC Corp., Changwon 641-846 (Korea, Republic of); Lim, K.; Baek, J.P. [SFA Engineering Corp., Asan 336-873 (Korea, Republic of)

    2016-11-01

    This paper describes both the test results of the bath type silver coating and the design of the bath to construct the silver coating plant for ITER thermal shield. The tests of small specimens made of SS304L and SS304LN were carried out to investigate the effect of the nitrogen content in SS304LN on the silver coating quality. The effect of different degreasing agents was also investigated to improve silver coating process. Small mock-up was tested to find a proper dipping direction during the electroplating process. Finally, noble bath design was conceived and structurally validated. Overall layout of silver coating plant is also shown in this paper.

  14. Dosimetric perturbations of a lead shield for surface and interstitial high-dose-rate brachytherapy

    International Nuclear Information System (INIS)

    Candela-Juan, Cristian; Granero, Domingo; Vijande, Javier; Ballester, Facundo; Perez-Calatayud, Jose; Rivard, Mark J

    2014-01-01

    In surface and interstitial high-dose-rate brachytherapy with either 60 Co, 192 Ir, or 169 Yb sources, some radiosensitive organs near the surface may be exposed to high absorbed doses. This may be reduced by covering the implants with a lead shield on the body surface, which results in dosimetric perturbations. Monte Carlo simulations in Geant4 were performed for the three radionuclides placed at a single dwell position. Four different shield thicknesses (0, 3, 6, and 10 mm) and three different source depths (0, 5, and 10 mm) in water were considered, with the lead shield placed at the phantom surface. Backscatter dose enhancement and transmission data were obtained for the lead shields. Results were corrected to account for a realistic clinical case with multiple dwell positions. The range of the high backscatter dose enhancement in water is 3 mm for 60 Co and 1 mm for both 192 Ir and 169 Yb. Transmission data for 60 Co and 192 Ir are smaller than those reported by Papagiannis et al (2008 Med. Phys. 35 4898–4906) for brachytherapy facility shielding; for 169 Yb, the difference is negligible. In conclusion, the backscatter overdose produced by the lead shield can be avoided by just adding a few millimetres of bolus. Transmission data provided in this work as a function of lead thickness can be used to estimate healthy organ equivalent dose saving. Use of a lead shield is justified. (paper)

  15. Remotely operated organic liquid waste incinerator for the fuels and materials examination facility

    International Nuclear Information System (INIS)

    Sales, W.L.; Barker, R.E.; Hershey, R.B.

    1980-01-01

    The search for a practical method for the disposal of small quantities of oraganic liquid waste, a waste product of metallographic sample preparation at the Fuels and Materials Examination Facility has led to the design of an incinerator/off-gas system to burn organic liquid wastes and selected organic solids. The incinerator is to be installed in a shielded inert-atmosphere cell, and will be remotely operated and maintained. The off-gas system is a wet-scrubber and filter system designed to release particulate-free off-gas to the FMEF Building Exhaust System

  16. 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)

  17. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  18. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  19. MMW [multimegawatt] shielding design and analysis

    International Nuclear Information System (INIS)

    Olson, A.P.

    1988-01-01

    Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

  20. Safety aspects and shield design of a Poton irradiator

    International Nuclear Information System (INIS)

    Mehta, S.K.; Nayak, A.R.; Bongirwar, D.R.; Modi, R.K.; Ramkumar, M.S.

    1998-01-01

    An irradiation plant, POTON, for irradiation of potatoes and onions is being set up at Nashik. Shield design and safety features of this plant incorporate some novel and innovative features like a compact cell, curved cell boundaries for smooth conveyor movement though the cell labyrinth and conform to ICRP and AERB design safety requirements. The safety features include multiple safety interlocks, audio-visual alarms, scram switches and trip wire for avoiding accidental exposures. (author)

  1. LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

    International Nuclear Information System (INIS)

    Kyllingstad, G.

    1977-01-01

    The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

  2. Comparison of calculational methods for liquid metal reactor shields

    International Nuclear Information System (INIS)

    Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

    1985-09-01

    A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

  3. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1994-02-01

    Concise description of progress in hot cell facility decommissioning at Risoe National Laboratory is presented. Removal of the large contaminated equipment has been completed, all the concrete cells have been finally cleaned. The total contamination left on the concrete walls is of the order of 1850 GBq. Preliminary smear tests proved the stack to be probably clean. The delay in project completion seems to be around 7 months, the remaining work being of rather conventional character. (EG)

  4. Decommissioning of the Risoe Hot Cell facility

    International Nuclear Information System (INIS)

    Carlsen, H.

    1991-08-01

    Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, removal and decontamination of large cell internals, and of large equipment such as glove boxes and steel boxes, in addition to dose commitments, are explained. Tables illustrating the analysis of smear tests, constants for contamination level examination, contamination and radiation levels after cleaning and total contamination versus measured radiation are included. (AB)

  5. A Survey of Structural Design of Diagnostic X-ray Imaging Facilities and Compliance to Shielding Design Goals in a Limited Resource Setting

    Directory of Open Access Journals (Sweden)

    Flavious B. Nkubli

    2017-11-01

    Full Text Available Purpose: To survey structural designs of x-ray rooms and compliance to shielding design goals of three x-ray imaging facilities. Methods and Materials: The survey was conducted in three radiodiagnostic centers in South East Nigeria, labeled X, Y and Z for anonymity. A stretchable non-elastic meter rule was used to measure x-ray room dimensions. A Vernier caliper was used to measure lead thickness while a calibrated digital survey meter Radalert 100x was used for radiation survey of controlled and uncontrolled areas. Simple statistical tools such as mean and standard deviation were used for analysis with the aid of Microsoft Excel version 2007. Results: Center X had a room dimension of 2.4 m × 2.1 m, Center Y had an x-ray room dimension of 3.6 m × 3.3 m, and Center Z had two x-ray rooms with identical dimensions of 6.3 m × 3.6 m. Measured exit radiation doses for controlled areas in all the centers were: 0.00152 mSv/wk; 0.00496 mSv/wk; 0.00168 mSv/wk; 0.00224 mSv/wk respectively. Lead was the common shielding material used. Conclusion: Based on the parameters studied, Center Z had the ideal room size and layout. Relative distances from the x-ray tubes to the nearest walls were not optimized in all the centers except in Center Z. Measured exit doses were within recommended limits except in Center Y. The location of the control consoles and measured doses were appropriate and within recommended design goals.

  6. 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)

  7. DWPF remotable television and cell lighting facilities

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

    The Defense Waste Processing Facility (DWPF) for radioactive waste vitrification at the Savannah River Plant (SRP) is now under construction. Development of specialized low cost television (TV) viewing equipment for in-cell and within-melter applications is now complete. High resolution TV cameras not originally designed for high radiation environments have been demonstrated in crane remotable packages to be well suited to the DWPF. High intensity in-cell lighting has also been demonstrated in crane remotable assemblies. These dual 1000 W units (2000 W total) are used to support the multiplicity of TV and cell window viewing requirements. 8 figures

  8. Determination of the neutron energy and spatial distributions of the neutron beam from the TSR-II in the large beam shield

    International Nuclear Information System (INIS)

    Clifford, C.E.; Muckenthaler, F.J.

    1976-01-01

    The TSR-II reactor of the ORNL Tower Shielding Facility has recently been relocated within a new, fixed shield. A principal feature of the new shield is a beam port of considerably larger area than that of its predecessor. The usable neutron flux has thereby been increased by a factor of approximately 200. The bare beam neutron spectrum behind the new shield has been experimentally determined over the energy range from 0.8 to 16 MeV. A high level of fission product gamma ray background prevented measurement of bare beam spectra below 0.8 MeV, however neutron spectra in the energy range from 8 keV to 1.4 MeV were obtained for two simple, calculable shielding configurations. Also measured in the present work were weighted integral flux distributions and fast neutron dose rates

  9. The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  10. Design and shielding calculation for a PET/CT facility

    International Nuclear Information System (INIS)

    Martin Escuela, J. M.; Palau San Pedro, A.; Lopez Diaz, A.

    2013-01-01

    Following the AAPM Task Group Report No. 108, the NCRP Report No. 147 recommendations and the Cuban's local regulations for nuclear medicine practice were carried out the safety planning and design of a new PET/CT facility for the Nuclear Medicine Department of 'Hermanos Ameijeiras' Hospital. It should be installed in the top floor of the NM building (3th floor), occupied by offices, classrooms and ancillaries areas, meanwhile in the second floor is working the conventional nuclear medicine department. The radiation doses were evaluated in areas of the second, third and quarter floor taking into account the pet isotope, the workload, the occupancy factors of each place, the use factors of different sources and the dose reduction factors, warranty the accomplish of the Cuban dose restrictions associated to the nuclear medicine practice. In each point of calculation was considered the contribution from each source to the total dose, as well as the contribution of the CT in the adjacent room to the imaging room. For the proper facility design was considered the transmission factors of the existing barriers, and calculated the new ones to be added between each source and the estimation point, keeping in mind the space limitations. The PET/CT design plan meet all the needs, the development of the project is consistent with the mission of the facility and the radiation protection regulations of nuclear medicine. (Author)

  11. HPGe detector shielding adjustment

    International Nuclear Information System (INIS)

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  12. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

    A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

  13. Hypervelocity Impact Performance of Open Cell Foam Core Sandwich Panel Structures

    Science.gov (United States)

    Ryan, S.; Ordonez, E.; Christiansen, E. L.; Lear, D. M.

    2010-01-01

    Open cell metallic foam core sandwich panel structures are of interest for application in spacecraft micrometeoroid and orbital debris shields due to their novel form and advantageous structural and thermal performance. Repeated shocking as a result of secondary impacts upon individual foam ligaments during the penetration process acts to raise the thermal state of impacting projectiles ; resulting in fragmentation, melting, and vaporization at lower velocities than with traditional shielding configurations (e.g. Whipple shield). In order to characterize the protective capability of these structures, an extensive experimental campaign was performed by the Johnson Space Center Hypervelocity Impact Technology Facility, the results of which are reported in this paper. Although not capable of competing against the protection levels achievable with leading heavy shields in use on modern high-risk vehicles (i.e. International Space Station modules), metallic foam core sandwich panels are shown to provide a substantial improvement over comparable structural panels and traditional low weight shielding alternatives such as honeycomb sandwich panels and metallic Whipple shields. A ballistic limit equation, generalized in terms of panel geometry, is derived and presented in a form suitable for application in risk assessment codes.

  14. Operational Radiation Protection in Synchrotron Light and Free Electron Laser Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Liu, James C.; Rokni, Sayed H.; /SLAC; Vylet, Vaclav; /Jefferson Lab

    2009-12-11

    The 3rd generation synchrotron radiation (SR) facilities are storage ring based facilities with many insertion devices and photon beamlines, and have low injection beam power (< few tens of watts), but extremely high stored beam power ({approx} 1 GW). The 4th generation x-ray free electron laser (FEL) facilities are based on an electron Linac with a long undulator and have high injection beam power (a few kW). Due to its electron and photon beam characteristics and modes of operation, storage ring and photon beamlines have unique safety aspects, which are the main subjects of this paper. The shielding design limits, operational modes, and beam losses are first reviewed. Shielding analysis (source terms and methodologies) and interlocked safety systems for storage ring and photon beamlines (including SR and gas bremsstrahlung) are described. Specific safety issues for storage ring top-off injection operation and FEL facilities are discussed. The operational safety program, e.g., operation authorization, commissioning, training, and radiation measurements, for SR facilities is also presented.

  15. Operational Radiation Protection in Synchrotron Light and Free Electron Laser Facilities

    International Nuclear Information System (INIS)

    Liu, James C.; Rokni, Sayed H.; Vylet, Vaclav

    2009-01-01

    The 3rd generation synchrotron radiation (SR) facilities are storage ring based facilities with many insertion devices and photon beamlines, and have low injection beam power (< few tens of watts), but extremely high stored beam power (∼ 1 GW). The 4th generation x-ray free electron laser (FEL) facilities are based on an electron Linac with a long undulator and have high injection beam power (a few kW). Due to its electron and photon beam characteristics and modes of operation, storage ring and photon beamlines have unique safety aspects, which are the main subjects of this paper. The shielding design limits, operational modes, and beam losses are first reviewed. Shielding analysis (source terms and methodologies) and interlocked safety systems for storage ring and photon beamlines (including SR and gas bremsstrahlung) are described. Specific safety issues for storage ring top-off injection operation and FEL facilities are discussed. The operational safety program, e.g., operation authorization, commissioning, training, and radiation measurements, for SR facilities is also presented.

  16. Qualification of academic facilities for small-scale automated manufacture of autologous cell-based products.

    Science.gov (United States)

    Hourd, Paul; Chandra, Amit; Alvey, David; Ginty, Patrick; McCall, Mark; Ratcliffe, Elizabeth; Rayment, Erin; Williams, David J

    2014-01-01

    Academic centers, hospitals and small companies, as typical development settings for UK regenerative medicine assets, are significant contributors to the development of autologous cell-based therapies. Often lacking the appropriate funding, quality assurance heritage or specialist regulatory expertise, qualifying aseptic cell processing facilities for GMP compliance is a significant challenge. The qualification of a new Cell Therapy Manufacturing Facility with automated processing capability, the first of its kind in a UK academic setting, provides a unique demonstrator for the qualification of small-scale, automated facilities for GMP-compliant manufacture of autologous cell-based products in these settings. This paper shares our experiences in qualifying the Cell Therapy Manufacturing Facility, focusing on our approach to streamlining the qualification effort, the challenges, project delays and inefficiencies we encountered, and the subsequent lessons learned.

  17. Method for dismantling shields

    International Nuclear Information System (INIS)

    Fukuzawa, Rokuro; Kondo, Nobuhiro; Kamiyama, Yoshinori; Kawasato, Ken; Hiraga, Tomoaki.

    1990-01-01

    The object of the present invention is to enable operators to dismantle shieldings contaminated by radioactivity easily and in a short period of time without danger of radiation exposure. A plurality of introduction pipes are embedded previously to the shielding walls of shielding members which contain a reactor core in a state where both ends of the introduction pipes are in communication with the outside. A wire saw is inserted into the introduction pipes to cut the shieldings upon dismantling. Then, shieldings can be dismantled easily in a short period of time with no radiation exposure to operator's. Further, according to the present invention, since the wire saw can be set easily and a large area can be cut at once, operation efficiency is improved. Further, since remote control is possible, cutting can be conducted in water and complicated places of the reactor. Biting upon starting the wire saw in the introduction pipe is reduced to facilitate startup for the rotation. (I.S.)

  18. Magnetic shield effect simulation of superconducting film shield covering directly coupled HTS dc-SQUID magnetometer

    International Nuclear Information System (INIS)

    Terauchi, N.; Noguchi, S.; Igarashi, H.

    2011-01-01

    A superconducting film shield over a SQUID ring improves the robustness of the SQUID with respect to magnetic noise. Supercurrent in the SQUID magnetometer and the superconducting film shield were simulated. The superconducting film shield reduces the influence of the external magnetic field on the SQUID ring. An HTS SQUID is a high sensitive magnetic sensor. In recent years, the HTS SQUID is widely used in various applications. In some applications, high robustness with respect to magnetic noise is required to realize stable operation at outside of a magnetic shielding room. The target of this paper is a directly coupled HTS dc-SQUID magnetometer. To enhance the robustness of the SQUID magnetometer, use of a superconducting thin film shield has been proposed. The magnetic field directly penetrating the SQUID ring causes the change of the critical current of Josephson junction, and then the SQUID magnetometer transitions into inoperative state. In order to confirm the magnetic shield effect of the superconducting film shield, electromagnetic field simulation with 3D edge finite element method was performed. To simulate the high temperature superconductor, E-J characteristics and c-axis anisotropy are considered. To evaluate the effect of the superconducting film shield, an external magnetic field which is supposed to be a magnetic noise is applied. From the simulation results, the time transition of the magnetic flux penetrating the SQUID ring is investigated and the effect of the superconducting film shield is confirmed. The amplitude of the magnetic flux penetrating the SQUID ring can be reduced to about one-sixth since the superconducting film shield prevents the magnetic noise from directly penetrating the SQUID ring.

  19. The New Anechoic Shielded Chambers Designed for Space and Commercial Applications at LIT

    Science.gov (United States)

    da Silva, Benjamim; Galvao, M. C.; Pereira, Clovis Solano

    2008-01-01

    The main objective of this paper is to present the capabilities of the new anechoic shielded rooms designed for space and commercial applications as part of the Integration and Testing Laboratory (LIT, Laboratorio de Integracao e Testes) in Brazil. A new anechoic shielded room named CBA2 has been in full operation since March 2007 and a remodeled chamber CBA1 is planned to be ready by the end of 2008, replacing an old facility which was in operation for the last 18 years. The Brazilian Space Program started with very small and simple satellites and the old CBA1 chamber was conceived in 1987 to accomplish the EMI/EMC tests not requiring significant volumes. Since the very beginning this facility was also used by the private sector for other applications mainly due to the absorption of digital electronics in all kind of products. The intense use of this facility during the last years, operating three shifts a day, caused a normal degradation and imposed several limitations. Therefore, a new totally remodeled chamber was designed considering the state of the art in terms of absorbers and associated instrumentation. On the other hand the facility CBA2 was conceived, designed and implemented to test large satellites taking into account the advance of the technology in terms of RF frequencies, power level, testing methodologies and several other factors. A very interesting and unique aspect of this project was the partnership between the private sector and governmental institution. As a result, the total investment was shared between several companies and consequently a time-sharing use of the facility as well.

  20. Passive magnetic shielding in MRI-Linac systems

    Science.gov (United States)

    Whelan, Brendan; Kolling, Stefan; Oborn, Brad M.; Keall, Paul

    2018-04-01

    Passive magnetic shielding refers to the use of ferromagnetic materials to redirect magnetic field lines away from vulnerable regions. An application of particular interest to the medical physics community is shielding in MRI systems, especially integrated MRI-linear accelerator (MRI-Linac) systems. In these systems, the goal is not only to minimize the magnetic field in some volume, but also to minimize the impact of the shield on the magnetic fields within the imaging volume of the MRI scanner. In this work, finite element modelling was used to assess the shielding of a side coupled 6 MV linac and resultant heterogeneity induced within the 30 cm diameter of spherical volume (DSV) of a novel 1 Tesla split bore MRI magnet. A number of different shield parameters were investigated; distance between shield and magnet, shield shape, shield thickness, shield length, openings in the shield, number of concentric layers, spacing between each layer, and shield material. Both the in-line and perpendicular MRI-Linac configurations were studied. By modifying the shield shape around the linac from the starting design of an open ended cylinder, the shielding effect was boosted by approximately 70% whilst the impact on the magnet was simultaneously reduced by approximately 10%. Openings in the shield for the RF port and beam exit were substantial sources of field leakage; however it was demonstrated that shielding could be added around these openings to compensate for this leakage. Layering multiple concentric shield shells was highly effective in the perpendicular configuration, but less so for the in-line configuration. Cautious use of high permeability materials such as Mu-metal can greatly increase the shielding performance in some scenarios. In the perpendicular configuration, magnetic shielding was more effective and the impact on the magnet lower compared with the in-line configuration.