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Sample records for bulk shielding facility

  1. Bulk shielding facility quarterly report, July, August, and September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, S. S.; Lance, E. D.; Thomas, J. R.

    1982-01-01

    The Bulk Shielding Reactor (BSR) operated at an average power level of 1919 kW for 85.74% of the time during July, August, and September. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The Pool Critical Assembly (PCA) was operated on three occasions for the Pressure Vessel Simulator Benchmark experiment.

  2. 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...

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

    International Nuclear Information System (INIS)

    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

  4. Technical specifications for the bulk shielding reactor

    International Nuclear Information System (INIS)

    This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

  5. Guidebook on radiation shielding safety for nuclear fuel facilities Q and A volume

    International Nuclear Information System (INIS)

    The Q and A volume of 'Guidebook on Radiation Shielding Safety for Nuclear Fuel Facilities' describes questions and answers which are commonly raised by the novices of shielding design and shielding safety evaluation. In this report, there are about 40 sets of Q and A which are classified by 7 different subjects, namely, (1) outlines of shielding, (2) methodology of shielding design, (3) shielding materials, (4) bulk shielding, (5) streaming, (6) skyshine, and (7) certification of shielding performance. The draft of the report has been discussed and summarized by the members of the specialists group for demonstration of shielding safety by analysis, committee for safety research on nuclear facilities. (author)

  6. New facility shield design criteria

    International Nuclear Information System (INIS)

    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. 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.

  8. Shielding design for positron emission tomography facility

    International Nuclear Information System (INIS)

    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

  9. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

  10. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

  11. Nuclear data relevant to shield design of FMIT facility

    International Nuclear Information System (INIS)

    Nuclear data requirements are reviewed for the design of the Fusion Materials Irradiation Test (FMIT) facility. This accelerator-based facility, now in the early stages of construction at Hanford, will provide high fluences in a fusion-like radiation environment for the testing of materials. The nuclear data base required encompasses the entire range of neutron energies from thermal to 50 MeV. In this review, we consider neutron source terms, cross sections for thermal and bulk shield design, and neutron activation for the facility

  12. Bulk shielding benchmark experiment at Frascati neutron generator (FNG)

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Santamarina, A.; Abidi, I.; Gastaldi, B.; Martini, M.; Marquette, J.P. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)

    1994-11-01

    In the framework of the European Fusion Technology Program, ENEA (Italian Agency for New Technologies, Energy and the Environment) - Frascati and CEA (Commissariat a` l`Energie Atomique) - Cadarache, in collaboration performed a bulk shielding benchmark experiment, using the 14-MeV Frascati neutron generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear data base and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The experimental results have been compared with numerical results calculated using both Sn and Monte Carlo transport codes and the cross section library EFF.1 (european fusion file).

  13. 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.

  14. The Tower Shielding Facility: Its glorious past

    International Nuclear Information System (INIS)

    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

  15. Benchmark calculations of target heat deposition and bulk shielding

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-09-01

    As a first step of a design study of the neutron science research center using an intense proton accelerator of 1.5 GeV with a current of 1 mA, a benchmark calculation was carried out with the NMTC/JAERI-MCNP-4A code system for the heat deposition in thick targets of Cu, Pb and U bombarded with 1.2 GeV protons. The thickness of bulk shielding around a spallation target was also estimated with the Moyer model and Sn calculation. It was found from these calculations that the code system reproduced well the experimental heat distribution around the beam axis. However, the code gave rather lower heat deposition at peripheral region of the target. As for the bulk shielding, it was estimated that the shielding made of iron having the thickness of 4 m surrounded by ordinary concrete with the thickness of 1 m was required for the 1.5 GeV proton incidence on a stopping-length Ta target with the diameter of 15 cm. (author)

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

    International Nuclear Information System (INIS)

    The charge state of 238U 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 238U beams are transported along the beam line. Secondary radiations generated by 238U 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. Radioisotope Power System Facility shielding analysis

    International Nuclear Information System (INIS)

    A series of calculations for the Radioisotope Power System Facility have been performed. These analyses have determined the shielding required for storage, testing, and transport of 238Pu heat source modules using the Monte Carlo code MCNP3B. The source terms and the assumptions used have been verified by comparison of calculated dose rates with measured ones. This paper describes the methodology used for shielding designs and the utilization of available variance reduction techniques to improve the computational efficiency. The new version of MCNP (MCNP3B) with a repeated structure capability was used. It decreased the chance for computer model errors and greatly decreased the model setup time. 2 refs., 3 figs., 2 tabs

  18. Operating manual for the Tower Shielding Facility

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  20. Early test facilities and analytic methods for radiation shielding: Proceedings

    International Nuclear Information System (INIS)

    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

  1. Early test facilities and analytic methods for radiation shielding: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T. (comp.) (Oak Ridge National Lab., TN (United States)); Ingersoll, J.K. (comp.) (Tec-Com, Knoxville, TN (United States))

    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.

  2. Radiation shielding facility and using method therefor

    International Nuclear Information System (INIS)

    A plurality of radiation shielding members are suspended and supported from a horizontally circular suspending beam by way of S-like suspending hooks. Wires are hung between the plurality of suspending fittings on the upper surface of the beam and a hook of a ceiling crane. The ceiling crane is an existent crane movably disposed to the ceiling in a building of a nuclear power plant containing a cylindrical vessel as a radiation source which is a member to be shielded. The radiation shielding member is a bag member formed by using a synthetic resin fabric or a rubber plate or a composite member thereof. A predetermined amount of a shielding material such as water is charged and kept in the bag member. The beam is suspended by the ceiling crane to transport the beam and each of the radiation shielding members altogether and lowered while being suspended so as to surround the outer circumference of the cylindrical vessel by each of the radiation shielding members. (I.N.)

  3. Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique.

    Science.gov (United States)

    Nasrabadi, M N; Mohammadi, A; Jalali, M

    2009-01-01

    In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required. PMID:19328700

  4. Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N. [Department of Nuclear Engineering, Faculty of Modern Sciences and Technologies, University of Isfahan, Isfahan 81746-73441 (Iran, Islamic Republic of)], E-mail: mnnasrabadi@ast.ui.ac.ir; Mohammadi, A. [Department of Physics, Payame Noor University (PNU), Kohandej, Isfahan (Iran, Islamic Republic of); Jalali, M. [Isfahan Nuclear Science and Technology Research Institute (NSTRT), Reactor and Accelerators Research and Development School, Atomic Energy Organization of Iran (Iran, Islamic Republic of)

    2009-07-15

    In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.

  5. Homogeneity test on heavy concrete shield wall for ACP facility

    International Nuclear Information System (INIS)

    The hot cell facility for research activities related to the electrolytic reduction of spent fuel, which is designed to permit a safe handling of radioactive materials up to 1,385 TBq, is scheduled to be constructed in 2005. The design features of the radiation safety are reviewed for the shield wall, rear door, shielding window, penetrations, toboggan, and the storage vault. The calculations by QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts and the gamma scanning test is described to examine the integrity of the shielding structure for the hot cell. The gamma scanning test is especially good at detecting any void and cracks in a heavy concrete wall and finding crevices between the wall and the devices frames. The shielding effectiveness and homogeneity of the hot cell wall, shield window, rear door etc., shall be measured by reading the activity level of the radiation

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

    International Nuclear Information System (INIS)

    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.

  7. 19 CFR 151.24 - Unlading facilities for bulk sugar.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 2 2010-04-01 2010-04-01 false Unlading facilities for bulk sugar. 151.24 Section... OF THE TREASURY (CONTINUED) EXAMINATION, SAMPLING, AND TESTING OF MERCHANDISE Sugars, Sirups, and Molasses § 151.24 Unlading facilities for bulk sugar. When dutiable sugar is to be imported in bulk, a...

  8. Radiation shielding for the Fermilab Vertical Cavity Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ginsburg, Camille; Rakhno, Igor; /Fermilab

    2010-03-01

    The results of radiation shielding studies for the vertical test cryostat VTS1 at Fermilab performed with the codes FISHPACT and MARS15 are presented and discussed. The analysis is focused on operations with two RF cavities in the cryostat. The vertical cavity test facility (VCTF) for superconducting RF cavities in Industrial Building 1 at Fermilab has been in operation since 2007. The facility currently consists of a single vertical test cryostat VTS1. Radiation shielding for VTS1 was designed for operations with single 9-cell 1.3 GHz cavities, and the shielding calculations were performed using a simplified model of field emission as the radiation source. The operations are proposed to be extended in such a way that two RF cavities will be in VTS1 at a time, one above the other, with tests for each cavity performed sequentially. In such a case the radiation emitted during the tests from the lower cavity can, in part, bypass the initially designed shielding which can lead to a higher dose in the building. Space for additional shielding, either internal or external to VTS1, is limited. Therefore, a re-evaluation of the radiation shielding was performed. An essential part of the present analysis is in using realistic models for cavity geometry and spatial, angular and energy distributions of field-emitted electrons inside the cavities. The calculations were performed with the computer codes FISHPACT and MARS15.

  9. Methodology for shielding design and evaluation in radiotherapy facilities

    International Nuclear Information System (INIS)

    The Government of the Republic of Cuba has decided to carry out a wide programme concerning the purchase of more than a dozen dual linear accelerators and, also; more than a dozen cobalt-60 units. Due to the lack of a national methodology for the design and calculation of shielding enclosures for radiotherapy units, the medical physicists from different hospitals began to use different methodologies, e.g. those in: a) Medical Physics Publishing. Shielding Techniques for Radiation Oncology Facilities. Patton H. McGinley. 1998.; b) National Council on Radiation Protection and Measurements, Structural shielding design and evaluation for medical use of X-rays and gamma-rays of energies up to 10 MeV, Report No. 49, NCRP, Washington, DC (1976).; c) National Council on Radiation Protection and Measurements, Radiation Protection Guidelines for 0.1 - 100 MeV Particle Accelerator Facilities, Report No. 51, NCRP, Washington, DC (1977). In some cases this caused the overestimation of the shielding thickness, when applying the values of dose constraints required by the Cuban regulations. The objective of the present work is to provide the medical physicists, the Radiation Safety Officers and other related professionals with a consistent methodology for the design and remodelation of bunkers hosting radiotherapy units but not using shielding doors. This work shows the validity of the above mentioned methodology, and the feasibility of designing door less bunkers for radiotherapy purposes. This methodology is considered to be self consistent and therefore no other complementary materials for its application are required. The experience so far confirms that; entry of realistic input data, and adequate application of sound engineering concepts when using this methodology leads to the achievement of enclosure shielding designs for radiotherapy units that comply with the dose constraints established by the Cuban regulations. Radiation shielding is attained having no over expenses on

  10. Activation measurements for the E.C. bulk shield benchmark experiment

    Science.gov (United States)

    Angelone, M.; Arpesella, C.; Martone, M.; Pillon, Mario

    1995-03-01

    The use of the absolute radiometric techniques for the E. C. bulk shield experiment at the 14 MeV Frascati Neutron Generator (FNG) is reported. In this application, the activity level, in some cases, results too low to be measured at the Frascati counting station. In these cases the radiometric measurements are performed using the low background HPGe detectors located at the underground laboratory of Gran Sasso d'Italia. The use of these detectors enhances the FNG capability of performing bulk shield benchmark experiments allowing the measurements of very low activation levels.

  11. Photon shielding calculations for a radiation waste facility benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Estes, G.P.; Urban, W.T.; Heath, A.R.

    1985-11-01

    Photon transport calculations have been performed for the ANS 6.2.1 radiation waste facility shielding benchmark using the continuous energy Monte Carlo code MCNP, and ONEDANT and TWODANT discrete ordinates codes. Comparisons are made of integral dose rates and flux spectra calculated with the three codes for various geometries, cross-section sets, and source and output energy group structures.

  12. Radiation Shielding Analysis of Electron Beam Accelerator Facility

    International Nuclear Information System (INIS)

    The objective of this technical report are to establish the radiation shielding technology of a high-energy electron accelerator to the facilities which utilize with electron beam. The technologies of electron beam irradiation(300 KeV -10 MeV) demand on the diverse areas of material processing, surface treatment, treatments on foods or food processing, improvement of metal properties, semiconductors, and ceramics, sterilization of medical goods and equipment, treatment and control of contamination and pollution, and so on. In order to acquire safety design for the protection of personnel from the radiations produced by electron beam accelerators, it is important to develop the radiation shielding analysis technology. The shielding analysis are carried out by which define source term, calculation modelling and computer calculations for 2 MeV and 10 MeV accelerators. And the shielding analysis for irradiation dump shield with 10 MeV accelerators are also performed by solving the complex 3-D geometry and long computer run time problem. The technology development of shielding analysis will be contributed to extend the further high energy accelerator development

  13. Benchmarking FENDL libraries through analysis of bulk shielding experiments on large SS316 assemblies for verification of ITER shielding characteristics

    International Nuclear Information System (INIS)

    FENDL-1 data base has been developed recently for use in ITER/EDA phase and other fusion-related design activities. It is now undergoing extensive testing and benchmarking using experimental data of differential and integral measured parameters obtained from fusion-oriented experiments. As part of co-operation between UCLA (U.S.) with JAERI (Japan) on executing the required neutronics R ampersand D tasks for ITER shield design, two bulk shielding experiments on large SS316 assemblies were selected for benchmarking FENDL/MG-1 multigroup data base and FENDL/MC-1 continous energy data base. The analyses with the multigroup data (performed with S8, P5, DORT calculations with shielded and unshielded data) also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo code was used by JAERI with the FENDL/MC-1 data. The results of this benchmarking is reported in this paper along with the observed deficiencies and discrepancies. 20 refs., 27 figs., 1 tab

  14. Effectiveness of Shield Termination Techniques Tested with TEM Cell and Bulk Current Injection

    Science.gov (United States)

    Bradley, Arthur T.; Hare, Richard J.

    2009-01-01

    This paper presents experimental results of the effectiveness of various shield termination techniques. Each termination technique is evaluated by two independent noise injection methods; transverse electromagnetic (TEM) cell operated from 3 MHz 400 MHz, and bulk current injection (BCI) operated from 50 kHz 400 MHz. Both single carrier and broadband injection tests were investigated. Recommendations as to how to achieve the best shield transfer impedance (i.e. reduced coupled noise) are made based on the empirical data. Finally, the noise injection techniques themselves are indirectly evaluated by comparing the results obtained from the TEM Cell to those from BCI.

  15. Guidebook on radiation shielding safety for nuclear fuel facilities practical volume

    International Nuclear Information System (INIS)

    The practical volume of 'Guidebook on Radiation Shielding Safety for Nuclear Fuel Facilities' is prepared as a report which describes practical designing procedures of shielding calculation for fuel cycle facilities. In this report, the facilities of uranium fuel fabrication, MOX fuel fabrication and fuel reprocessing, and a fuel transport cask are taken up as typical facilities in the fuel cycle. The practical procedures for these facilities have been divided by four subjects, namely, (1)practical methodology of shielding design, (2)procedures of shielding calculation, (3)examples of shielding calculations and (4)check sheet of shielding calculation. The draft of the report has been discussed and summarized by the members of 'Working group on methodology of shielding safety' chaired by Dr.T.Kosako of the University of Tokyo. The working group belongs to the specialists group for demonstration of shielding safety by analysis, committee for safety research on nuclear facilities. (author)

  16. The bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Angelone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Martone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Pillon, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Santamarina, A. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Abidi, I. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Gastaldi, B. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Martini, M. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Marquette, J.P. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France)

    1995-03-01

    In the design of next-step fusion devices such as NET/ITER the nuclear performance of shielding blankets is of key importance in terms of nuclear heating of superconducting magnets and radiation damage. In the framework of the European Fusion Technology Program, ENEA Frascati and CEA Cadarache in collaboration performed a bulk shielding benchmark experiment using the 14MeV Frascati Neutron Generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear database and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14MeV neutrons. The neutron reaction rates at various depths inside the block have been measured using fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S{sub n} and Monte Carlo transport codes and the cross-section library EFF.1 (European Fusion File). (orig.).

  17. Magnetic shielding of an inhomogeneous magnetic field source by a bulk superconducting tube

    International Nuclear Information System (INIS)

    Bulk type-II irreversible superconductors can act as excellent passive magnetic shields, with a strong attenuation of low frequency magnetic fields. Up to now, the performances of superconducting magnetic shields have mainly been studied in a homogenous magnetic field, considering only immunity problems, i.e. when the field is applied outside the tube and the inner field should ideally be zero. In this paper, we aim to investigate experimentally and numerically the magnetic response of a high-Tc bulk superconducting hollow cylinder at 77 K in an emission problem, i.e. when subjected to the non-uniform magnetic field generated by a source coil placed inside the tube. A bespoke 3D mapping system coupled with a three-axis Hall probe is used to measure the magnetic flux density distribution outside the superconducting magnetic shield. A finite element model is developed to understand how the magnetic field penetrates into the superconductor and how the induced superconducting shielding currents flow inside the shield in the case where the emitting coil is placed coaxially inside the tube. The finite element modelling is found to be in excellent agreement with the experimental data. Results show that a concentration of the magnetic flux lines occurs between the emitting coil and the superconducting screen. This effect is observed both with the modelling and the experiment. In the case of a long tube, we show that the main features of the field penetration in the superconducting walls can be reproduced with a simple analytical 1D model. This model is used to estimate the maximum flux density of the emitting coil that can be shielded by the superconductor. (paper)

  18. Shielding design for a proton medical accelerator facility

    International Nuclear Information System (INIS)

    Source terms and attenuation lengths for neutrons produced by 250 MeV protons on iron, copper and soft tissue, calculated with the FLUKA Monte Carlo code, were used for the shielding calculations (walls, ceilings, and floors) for the National Centre for Oncological Hadrontherapy to be built in Italy. Appropriate hypotheses on the proton current, beam loss factors, duty factors, occupancy factors and use factors of the shields were adopted. A dose equivalent limit of 1 mSv per year in the areas where the public has access and of 2 mSv per year for facility personnel were assumed. Shielding requirements vary from 1.5 m to about 4 m of ordinary concrete. The results agree with Monte Carlo simulations of the complete geometry of the facility obtained in a previous work. The access mazes to the treatment rooms were designed by the LCS Monte Carlo code by optimizing the length and section of their legs and their wall thicknesses with the dose equivalent limit of 2 mSv per year, fixed in the areas accessed by personnel. The resulting annual neutron dose equivalent at the maze mouth is 0.6 mSv

  19. Shielding aspects of accelerators, targets and irradiation facilities

    International Nuclear Information System (INIS)

    Particle accelerators have evolved over the last half-century 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: increased particle beams intensities, higher flexibility, etc., giving rise to new radiation shielding aspects and problems. These Proceedings review newer accelerator facilities, identify problematic aspects concerning radiation shielding that need to be solved, and indicate areas where international co-operation and co-ordination are highly desirable. (authors). 480 refs., 200 figs., 48 tabs

  20. Bulk shielding facility quarterly report, October, November, and December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, III, S. S.; Lance, E. D.; Thomas, J. R.

    1977-08-01

    The BSR operated at an average power level of 1,836 kw for 78.01 percent of the time during October, November, and December. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The PCA was used in training programs and was operated on two occasions when the University of Kentucky students actively participated in training laboratories.

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

    International Nuclear Information System (INIS)

    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

  2. Individual patient shielding for a proton eye therapy facility

    International Nuclear Information System (INIS)

    The first proton ocular radiotherapy facility in Poland has been developed at the Institute of Nuclear Physics (IFJ PAN), Krakow, in collaboration with the Clinic of Ophthalmology and Ocular Oncology of the Collegium Medicum of the Jagiellonian University in Krakow and the Centre of Oncology of the Maria Sklodowska-Curie Memorial Institute Krakow Branch. For the proton beam from the AIC-144 isochronous cyclotron the unwanted patient dose due to secondary radiation - predominantly neutrons - has been evaluated using the MCNPX code. Optimisation of the beam forming elements and designing of the additional patient shielding has been performed to minimize the unwanted patient's dose.

  3. Radiation shielding for the ITER neutral beam test facility

    International Nuclear Information System (INIS)

    The NB system for the International Thermonuclear Experimental Reactor (ITER) consists of two heating and current drive (H and CD) NB injectors and a diagnostic neutral beam (DNB) injector. The NB accelerates negative deuterium ions with maximum energy of 1 MeV and maximum beam current of 40 A. The ITER (H and CD) NB will be tested in the Neutral Beam Test Facility (NBTF) that will be located in Italy, near Padua. The performance test will be based on different operation phases starting with low energy hydrogen beam. In the initial testing phase for many months the machine will operate with hydrogen only and with deuteron at a reduced intensity suggesting the possibility of hosting the device in a light shielding room/area. In the paper the study performed to evaluate the minimum shielding needed in connection with the different operation phases is shown. The source terms were calculated starting from neutron source characterisation and then assessing article transport in the ITER NB structure with a mathematical model of the components geometry that was implemented into MCNP computer code. The neutron source definition was outlined considering both D-D and D-T neutron production. Shielding was assessed for hydrogen operation only and for 20, 60, 100 and 1000 kV (full energy) deuteron acceleration, accounting for the associated beam current intensity. Related results are presented and discussed in the paper. (author)

  4. Training Software for the Bulk Handling Facility

    International Nuclear Information System (INIS)

    In 2013, the International Atomic Energy Agency, Department of Safeguards, applied safeguards in 180 States with safeguards agreements in force, with implementation of safeguards at over 600 facilities. To support the Department of Safeguards in fulfiling its mission, the training section holds over 100 training courses yearly to help inspectors and analysts develop the necessary knowledge, skills and abilities. An effective training programme must be able to adapt and respond to changing organizational training needs. Virtual training technologies have the potential to broaden the spectrum of possible training activities, enhance the effectiveness of existing courses, optimize off-site training and activities, and possibly increase trainee motivation and accelerate learning. Ultimately, training is about preparation - being ready to perform in different environments, under a range of conditions or unknown situations. Virtual environments provide this opportunity for the trainee to encounter and train under different scenarios not possible in real facilities. This paper describes the training software developed for fuel fabrication facilities to be used by both national inspectors and IAEA inspectors. The model includes interactive modules to explain each of the six main fuel fabrication processes. It also includes verification instruments at specific locations with animations that illustrate how to operate the instrument, verify the material and report. Additionally, the software integrates an evaluation mode to allow the trainee and the instructor to track progress and evaluate learning. Overall, the model can be used for individual training, or integrated into a training course where the instructor can draw on the virtual model to enhance the overall effectiveness of the training. (author)

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

    International Nuclear Information System (INIS)

    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

  6. High level process shielded line (CBP) and high level analysis shielded line (CBA): two of the newest facilities of ATALANTE facility

    International Nuclear Information System (INIS)

    The two newest facilities in the ATALANTE complex, a high-level shielded process line (CBP) and high-level shielded analysis line (CBA), are described and their work programs detailed, notably the dissolution in CBP of 15 kg of spent fuel to demonstrate the technological feasibility of partitioning the minor actinides. The analytical support role of CBA is also discussed. (authors)

  7. Coupled Monte Carlo - Discrete ordinates computational scheme for three-dimensional shielding calculations of large and complex nuclear facilities

    International Nuclear Information System (INIS)

    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 simulation technique. This work proposes a dedicated computational approach for coupled Monte Carlo - deterministic 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. To enable the coupling of these two different computational methods, a mapping approach has been developed for calculating the discrete ordinates angular flux distribution from the scored data of the Monte Carlo particle tracks crossing a specified surface. The approach has been implemented in an interface program and validated by means of test calculations using a simplified three-dimensional geometric model. Satisfactory agreement was obtained for the angular fluxes calculated by the mapping approach using the MCNP code for the Monte Carlo calculations and direct three-dimensional discrete ordinates calculations using the TORT code. In the next step, a complete program system has been developed for coupled three-dimensional Monte Carlo deterministic transport calculations by integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and the mapping interface program. Test calculations with two simple models have been performed to validate the program system by means of comparison calculations using the

  8. Upgrading the Neutron Radiography Facility in South Africa (SANRAD): Concrete Shielding Design Characteristics

    Science.gov (United States)

    de Beer, F. C.; Radebe, M. J.; Schillinger, B.; Nshimirimana, R.; Ramushu, M. A.; Modise, T.

    A common denominator of all neutron radiography (NRAD) facilities worldwide is that the perimeter of the experimental chamber of the facility is a radiation shielding structure which,in some cases, also includes flight tube and filter chamber structures. These chambers are normally both located on the beam port floor outside the biological shielding of the neutron source. The main function of the NRAD-shielding structure isto maintain a radiological safe working environment in the entire beam hall according to standards set by individual national radiological safety regulations. In addition, the shielding's integrity and capability should not allow, during NRAD operations, an increase in radiation levels in the beam port hall and thus negatively affectadjacent scientific facilities (e.g. neutron diffraction facilities).As a bonus, the shielding for the NRAD facility should also prevent radiation scattering towards the detector plane and doing so, thus increase thecapability of obtaining better quantitative results. This paper addresses Monte Carlo neutron-particletransport simulations to theoretically optimize the shielding capabilities of the biological barrierfor the SANRAD facility at the SAFARI-1 nuclear research reactor in South Africa. The experimental process to develop the shielding, based on the principles of the ANTARES facility, is described. After casting, the homogeneity distribution of these concrete mix materials is found to be near perfect and first order experimental radiation shielding characteristicsthrough film badge (TLD) exposure show acceptable values and trends in neutron- and gamma-ray attenuation.

  9. Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    张志宏; 夏晓彬; 蔡军; 王建华; 李长园; 葛良全; 张庆贤

    2015-01-01

    The Chinese Academy of Science has launched a thorium-based molten-salt reactor (TMSR) research project with a mission to research and develop a fission energy system of the fourth generation. The TMSR project intends to construct a liquid fuel molten-salt reactor (TMSR-LF), which uses fluoride salt as both the fuel and coolant, and a solid fuel molten-salt reactor (TMSR-SF), which uses fluoride salt as coolant and TRISO fuel. An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle. Preliminary conceptual shielding design has also been performed to develop bulk shielding. In this study, the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics (CFD) analysis. The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91 µSv/h in the radial direction, 1.16 µSv/h above and 1.33 µSv/h below the bulk shielding. All the radiation dose rates due to the core were below the design criteria. Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K, which was below the required limit value. The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

  10. The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

    International Nuclear Information System (INIS)

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both SN and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

  11. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Angelone, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Martone, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Petrizzi, L. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Pillon, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Rado, V. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Santamarina, A. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex (France)); Abidi, I. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex (France)); Gastaldi, G. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex

    1994-09-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The [gamma]-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

  12. Benchmark experiment on stainless steel bulk shielding at Frascati neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V. [ENEA, Frascati (Italy). Centro Ricerche Energia - Area Energia e Innovazione

    1994-11-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L`Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S{sub N} and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the (ENEA Italian Agency for New Technologies, Energy and Environment) team.

  13. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    Science.gov (United States)

    Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J. P.; Martini, M.

    1994-09-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat à l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File).

  14. Shielding experiments

    International Nuclear Information System (INIS)

    Shielding mock-up experiments for Prototype Fast Breeder Reactor (PFBR) and Advanced Heavy Water Reactor (AHWR) are carried out in shielding corner facility of APSARA reactor, to assess the overall accuracy of the codes and nuclear data used in reactor shield design. As APSARA is a swimming pool-type thermal reactor, for fast reactor experiments, typical fast reactor shielding facility was created by using uranium assemblies as spectrum converter. The flux was also enhanced by replacing water by air. Experiments have been carried out to study neutron attenuation through typical fast reactor radial and axial bulk shielding materials such as steel, sodium, graphite, borated graphite and boron carbide. A large number of reaction rates, sensitive to different regions of the neutron energy spectrum, were measured using foil activation and Solid State Nuclear Track Detector (SSNTD) techniques. These experimental results were analysed using computational tools normally used in design calculations, viz., discrete ordinate transport codes with multigroup cross section sets. Comparison of measured reaction rates with calculations provided suitable bias factors for parameters relevant to shield design, such as sodium activation, fast neutron fluence, fission equivalent fluxes etc. The measured neutron spectrum on the incident face of shield model compares well with the calculated fast reactor blanket leakage neutron spectrum. The comparison of calculated reaction rates within shield model indicate that the calculations suffer from considerable uncertainties, in shield models with boron carbide/borated graphite. For AHWR shielding experiments, no spectrum converter was used as it is also a thermal reactor. Radiation streaming studies through penetrations/ducts of various shapes and sizes relevant to AHWR shielding were carried out. (author)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  17. Fire Events Effect on Concrete Shielding of 60Co Industrial Irradiation Facilities

    International Nuclear Information System (INIS)

    Concrete is the most important structural which is used in radiation shielding for owing to its low price and good shielding performance. In concrete structures, of industrial irradiation facilities cracks occur due to thermal stress, hydration he at, weather, dry storage radiation of sources and fire events nuclear reaction. Shielding performance is effected according to the crack size geometry (width, depth and bending). However, there are no design criteria providing the allowable crack size limits for concrete shielding. Three different methods are applied in this work wbere the measuring instrument for qualifying the effect of cracks on shielding performance in concrete mass. The correlation between crack size and concrete shielding performance is deduced. The surface dose rate increased logarithmically according to the increase in crack width. The results are compared by the effects of fire events inside industrial irradiation facilities cold sterilization which led to change the effected mechanical parts and electrical components. Radiation concrete shielding is repaired. Radiation standard measurements outside irradiation concrete facility recorded normal radiation values. Radiation safety is maintained

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  20. Calculation of shielding of X rays in radiotherapy facilities with computer aid

    International Nuclear Information System (INIS)

    This work presents a methodology for calculation of shielding of X rays in radiotherapy facilities with computer aid. A friendly program, called RadTeraX, was developed in programming language Delphi that, through manual data input of a basic project of architecture and of some parameters, interprets the geometry and calculates the shielding of the walls, ground and roof of a radiotherapy installation for X rays. As a final product, this program supplies a graphic screen in the computer with all the input data and the calculation of the shielding, besides the respective calculation memory. Still today, in Brazil, the calculation of the shielding for radiotherapy facilities with X rays has been made based on recommendations of NCRP-49, that establishes a necessary calculation methodology to the elaboration of a shielding project. However, in high energies, where it is necessary the construction of a maze, NCRP-49 is insufficient, so that in this field, studies were made originating an article that proposes a solution for the problem and this solution was implemented in the program. The program can be applied in the practical execution of shielding projects for radiotherapy facilities and in didactic way in comparison with NCRP-49 and has been registered under number 00059420 at INPI - Instituto Nacional da Propriedade Industrial (National Institute of Industrial Property). (author)

  1. Early Test Facilities and Analytic Methods for Radiation Shielding

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    1992-01-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting held in Chicago, Illinois on November 15 20,1992. The meeting is of special significance since it commemorates the 50th anniversary of the first controlled nuclear chain reaction, which occurred, not coincidentally, in Chicago. 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.

  2. Preliminary shielding estimates for the proposed National ISOL Radioactive Ion Beam (RIB) Facility at Oak Ridge

    International Nuclear Information System (INIS)

    ORNL built a first-generation Radioactive Ion Beam (RIB) facility for astrophysics and nuclear physics research; it was named Holifield Radioactive Ion Beam Facility (HRIBF) and is based on the Isotope Separator On Line (ISOL) technique. Planning is underway for a second- generation facility, the National ISOL RIB facility at Oak Ridge; it will build on the existing HRIBF and may utilize many existing components and shielded areas. Preliminary upgrade plan for the new facility includes: adding a superconducting booster for the tandem accelerator; replacing the 1960-vintage, 60-MeV proton, 50-microamp ORIC (Oak Ridge Isochronous Cyclotron) with a modern 200-MeV proton, 200-microamp cyclotron; and building a high-power 238U fission target to accept the 200-MeV proton beam. This report summarizes the results of a preliminary 1-D shielding analysis of the proposed upgrade, to determine the shielding requirements for a 0.25 mrem/h dose rate at the external surface of the exclusion area. Steel shielding weights ranging from 60 to 100 metric tons, were considered manageable; these could be reduced by a factor of 2 to 3 if the orientation of the proposed target station was changed

  3. Applications of point kernel estimates to the fuel conditioning facility shield test program

    International Nuclear Information System (INIS)

    Use of a multigroup point kernel gamma ray attenuation program helped Argonne National Laboratory complete a project that verified the integrity of the shields in the Fuel Conditioning Facility (FCF). Test procedures were developed based on predictions of dose equivalents as functions of source strengths, source-to-detector distances, and thickness of various shield materials. Tables and plots of these data were used to select and position the test sources and to compare as-built shields to their design thicknesses. Part of the program involved a study of the penetration of photons from spent fuel as a function of cooling time. Such information is important to estimate the effectiveness of FCF shields on mixed fission product sources from various reactors

  4. Verification of shielding calculation on the DIII-D facility at La Jolla, California

    International Nuclear Information System (INIS)

    Shielding calculations were performed for the DIII-D facility at La Jolla to independently assess the biological dose from radiation emitted during operation. These calculations for both the fully shielded and bare configurations are in essential agreement with those done by Gulf. In addition to the basic test problems run by Gulf, a bare configuration with additional air outside the facility area was calculated. The addition of air to the bare configuration caused the dose at 100 meters from the DIII-D center-line to increase by fifty five percent. The inclusion of the various elemental constituents in the soil composition may change the calculated dose, but will not change the shielding factor nor invalidate the overall conclusion of this report. The overall conclusion is that Gulf and LLNL results are in general agreement. 5 refs., 11 figs., 5 tabs

  5. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  6. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

  7. Neutron Shielding Design for 4π BaF2 Detector Facility

    Institute of Scientific and Technical Information of China (English)

    HUANG; Xing; ZHANG; Qi-wei; HE; Guo-zhu; CHENG; Pin-jing; TANG; Hong-qing; ZHOU; Zu-ying

    2013-01-01

    Neutrons within energy range of 5 to 300 keV can be produced by pulsed proton beam striking thick lithium target,based on the HI-13 tandem accelerator.Neutron shielding is necessary when the Gamma-ray Total Absorption Facility(GTAF)is applied to measured(n,γ)reaction cross sections

  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)

    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. Efficient time-independent method for conceptual design optimization of the national ignition facility primary shield

    International Nuclear Information System (INIS)

    Minimum-cost design concepts of the primary shield for the National (laser fusion) Ignition Facility are sought with the help of the SWAN optimization code. 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 the time-dependent problem to be addressed using time-independent transport calculations, thus significantly simplifying and accelerating the design process. The search for constituents that will minimize the shield cost is guided by the newly defined equal cost replacement effectiveness functions. The minimum-cost shield design concept consists of a mixture of polyethylene and low-cost, low-activation materials, such as CaCO3 or silicon carbide, with boron added near the shield boundaries. An alternative approach to the target chamber design is analyzed. It involves placing the shield interior, rather than exterior to the main aluminum structural wall of the target chamber. The resulting inner shield design approach was found to be more expensive but inherently safer; the overall inventory of radioactive activation product it contains is one to two orders of magnitude lower than in the conventional design approach. 21 refs., 16 figs., 15 tabs

  11. Design and shielding calculation for a PET/CT facility

    International Nuclear Information System (INIS)

    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)

  12. Dose conversion coefficients in the shielding design calculation for high energy proton accelerator facilities

    International Nuclear Information System (INIS)

    Dose quantity in the shielding design calculation was changed from the 1 cm depth dose equivalent to effective dose on the occasion of the introduction of the International Commission on Radiological Protection (ICRP) 1990 Recommendations (ICRP Publication 60) into domestic laws. As dose conversion coefficients in the shielding design calculation for accelerator facilities, the values for front irradiation (AP irradiation geometry) of neutrons below 20 MeV based on the ICRP Publication 74 are listed in the accompanying table of the domestic laws, but the values for neutrons above 20 MeV are not shown in the accompanying table. The status of dose conversion coefficients for neutrons above 20 MeV was surveyed and the effective dose rates behind the concrete shield of proton accelerator facilities were obtained by using typical neutron spectra and various dose conversion coefficients. As a result of consideration, the effective dose conversion coefficients for front irradiation of neutrons above 20 MeV evaluated by using HERMES code system was recommended for high energy neutrons in the shielding design calculation of proton accelerator facilities and 77 energy group averaged dose conversion coefficients was produced from thermal energy to 2 GeV. (author)

  13. Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

    International Nuclear Information System (INIS)

    This work is concerned with the study of the distribution attenuation of doses of thermal neutrons emitted directly from the core of research reactor in ordinary concrete shield. In practice it is not possible to identify the reactor thermal neutrons in the emitted continuos neutron spectrum. Therefore, measurement was carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li2B4O7:Mn thermoluminescent dosimeters. The data obtained were analyzed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc-collimated beam and infinite plane mono-directional sources

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

    Energy Technology Data Exchange (ETDEWEB)

    Michael Kruzic

    2007-09-01

    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&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.

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

    International Nuclear Information System (INIS)

    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

  16. Improved Methodology Application for 12-Rad Analysis in a Shielded Facility at SRS

    International Nuclear Information System (INIS)

    The DOE Order 420.1 requires establishing 12-rad evacuation zone boundaries and installing Criticality Accident Alarm System (CAAS) per ANS-8.3 standard for facilities having a probability of criticality greater than 10-6 per year. The H-Canyon at the Savannah River Site (SRS) is one of the reprocessing facilities where SRS reactor fuels, research reactor fuels, and other fissile materials are processed and purified using a modified Purex process called H-Modified or HM Process. This paper discusses an improved methodology for 12-rad zone analysis and its implementation within this large shielded facility that has a large variety of criticality sources and scenarios

  17. Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

    2008-10-31

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured

  18. An Evaluation on the Radiation Shielding of the Radwaste Drum Assay Facility

    International Nuclear Information System (INIS)

    In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

  19. Radiation shielding for superconducting RF cavity test facility at A0

    International Nuclear Information System (INIS)

    The results of Monte Carlo radiation shielding study performed with the MARS15 code for the vertical test facility at the A0 north cave enclosure at Fermilab are presented and discussed. The vertical test facility at the A0 north cave is planned to be used for testing 1.3 GHz single-cell superconducting RF cavities with accelerating length of 0.115 m. The operations will be focused on high accelerating gradients--up to 50 MV/m. In such a case the facility can be a strong radiation source (1). When performing a radiation shielding design for the facility one has to take into account gammas generated due to interactions of accelerated electrons with cavity walls and surroundings (for example, range of 3.7-MeV electrons in niobium is approximately 3.1 mm while the thickness of the niobium walls of such RF cavities is about 2.8 mm). The electrons are usually the result of contamination in the cavity. The radiation shielding study was performed with the MARS15 Monte Carlo code (2). A realistic model of the source term has been used that describes spatial, energy and angular distributions of the field-emitted electrons inside the RF cavities. The results of the calculations are normalized using the existing experimental data on measured dose rate in the vicinity of such RF cavities

  20. Radiation shielding for superconducting RF cavity test facility at A0

    Energy Technology Data Exchange (ETDEWEB)

    Dhanaraj, N.; Ginsburg, C.; Rakhno, I.; Wu, G.; /Fermilab

    2008-11-01

    The results of Monte Carlo radiation shielding study performed with the MARS15 code for the vertical test facility at the A0 north cave enclosure at Fermilab are presented and discussed. The vertical test facility at the A0 north cave is planned to be used for testing 1.3 GHz single-cell superconducting RF cavities with accelerating length of 0.115 m. The operations will be focused on high accelerating gradients--up to 50 MV/m. In such a case the facility can be a strong radiation source [1]. When performing a radiation shielding design for the facility one has to take into account gammas generated due to interactions of accelerated electrons with cavity walls and surroundings (for example, range of 3.7-MeV electrons in niobium is approximately 3.1 mm while the thickness of the niobium walls of such RF cavities is about 2.8 mm). The electrons are usually the result of contamination in the cavity. The radiation shielding study was performed with the MARS15 Monte Carlo code [2]. A realistic model of the source term has been used that describes spatial, energy and angular distributions of the field-emitted electrons inside the RF cavities. The results of the calculations are normalized using the existing experimental data on measured dose rate in the vicinity of such RF cavities.

  1. A decade of radiological and shielding experience at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF) was designed to permit irradiation testing of fuels and materials to support the commercial development of liquid-metal-cooled fast reactors (LMRs). A secondary objective was to gain experience in the design, construction, and operation of a relatively large LMR. The radiological experience gained from the operation of the facility as it applies to the area of radiation protection and shielding is presented. Experience from 8 yr of FFTF operation has demonstrated that radiological safety can be achieved in large LMRs. Layout of plant equipment in shielded compartments, careful operational planning, and adherence to procedures have combined to minimize personnel doses at FFTF and the release of radioactivity to the environment. The experience derived form the design, construction, and operation of FFTF should be of inestimable value in supporting future LMR development

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

    International Nuclear Information System (INIS)

    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. Bulk shielding experiments on large SS316 assemblies bombarded by D-T neutrons. Volume I: experiment

    International Nuclear Information System (INIS)

    SS316 is one of the most promising candidates for the shielding and structural material of next fusion devices such as ITER. Benchmark experiments to examine the bulk shielding performance of SS316 for D-T neutrons, particularly deep penetration, were performed by using the D-T neutron source FNS in Japan Atomic Energy Research Institute as the '94 ITER/EDA task (T-16). This report compiles the experimental system, measuring procedures and the measured data. The analysis of the experiment is described separately in the Volume II. The test region of the experimental assembly was a cylindrical SS316 of 1200 mm in diameter and 1118 mm in thickness which was located at 300 mm from the D-T neutron source (Assembly no.1). A source reflector of 200 mm-thick SS316 surrounding the D-T neutron source was added to the assembly no.1 to simulate a neutron field of a fusion reactor (Assembly no.2). The measured data for i) neutron spectra in energy regions of MeV, keV and eV, ii) neutron activation reaction rates, iii) fission rates, iv) gamma-ray spectra and v) gamma-ray heating rates were obtained from the test region surface to the depth of 914 mm in the test region. The consistency of the measured data and the effect of the source reflector were examined from the comparison among the measured data. (author) 51 refs

  4. Shielding Design for Adjacent, Underground Buildings of a Megavoltage Radiotherapy Facility.

    Science.gov (United States)

    Sanz, Darío Esteban

    2016-07-01

    In a radiotherapy facility, safety in areas next to the treatment room can be of concern when irradiating downward due to oblique x-ray transmission through the floor and/or walls, especially in areas immediately adjacent or underground. Even when there is no basement underneath, a usual conservative solution is to build a thick concrete slab as the base for the treatment room. Of course, this implies deeper soil excavation and higher associated costs. As a convenient alternative, the limiting walls can be buried a certain depth below floor level to shield oblique, downward irradiation. Besides, for space considerations, laminated barriers are usually employed, and some additional shielding to the floor may be required (L-shaped barriers). In this work, the author introduces an analytical method for calculating the required wall penetration below floor level or, alternatively, the additional floor shielding for L-shaped barriers, taking into account in either case the attenuation properties of the earth underneath the vault. Interestingly, the required penetration depth for a given wall barrier (primary or secondary), relative to a reference thickness, is only a function of basic attenuation data. Likewise, for a laminated, lead-concrete barrier, the required dimensions depend on the relative amount of lead used for the wall and on the corresponding attenuation data. The shielding design criteria developed in this work to protect underground nearby sites is conservative in nature, yet it yields optimal shield dimensions for wall footing and for wall-floor shielding, avoiding the need to construct oversized concrete slab floors. PMID:27218288

  5. 40 CFR 63.11086 - What requirements must I meet if my facility is a bulk gasoline plant?

    Science.gov (United States)

    2010-07-01

    ... facility is a bulk gasoline plant? 63.11086 Section 63.11086 Protection of Environment ENVIRONMENTAL... Source Category: Gasoline Distribution Bulk Terminals, Bulk Plants, and Pipeline Facilities Emission... gasoline plant? Each owner or operator of an affected bulk gasoline plant, as defined in § 63.11100,...

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10B (n, α) 7Li 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)

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

    International Nuclear Information System (INIS)

    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

  9. 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

  10. Design of special lead shielding facilities for medium and low energy gamma radiation

    International Nuclear Information System (INIS)

    The cardinal principles of radiation protection from external sources are based on four factors: time, distance, shielding and activity. These factors are more or less rigorously observed inside the hot room of the nuclear medicine laboratory. Unfortunately the importance of radiation protection during data acquisition, storage of radioisotopes and waste products are often overlooked. The patients are a source of significant radiation and the nuclear medicine personnel must consciously take measures for protection against this source during patient handling. There are many commercial shielding materials for the partition from radioisotopes. Commercially available lead cupboard and lead glass used as partition from radioisotopes are very expensive. A project was therefore undertaken to develop practical low cost shielding against radiation sources. The present article outlines the various lead shielding facilities designed and built for medium and low energy gamma radiation. The designs were at first sketched out on paper giving specific shapes and measurement of each structure. Model structures were then made accordingly and the protective capacities of these structures were checked by mathematical calculation from the equation of gamma ray attenuation. In the design and structures lead plating thickness was in between 0.30 to 5.0 cm. Correction and restructuring of the models were undertaken to achieve the designs satisfactory. Different structures served different aspects of radiation protection. These structures sculptured as per designs are now in use for radiation protection at the Institute of Nuclear Medicine. (author) 2 tabs., 6 figs., 4 refs

  11. Construction of a Post-Irradiated Fuel Examination Shielded Enclosure Facility

    International Nuclear Information System (INIS)

    construction activities, which will include facility modifications and construction of one shielded enclosure. Follow-up activities will be to construct two additional shielded enclosures to complete the suite of three separate but connected remote operated examination areas. Equipment purchases are to be capital procurement spread out over several years on a funded schedule. This paper discusses safety and operational considerations given during the conceptual design phase of the project. The paper considers such things as project material at risk (MAR), new processes and equipment, potential hazards, and the major modification evaluation process to determine if a preliminary Documented Safety Analysis (PDSA) is required. As part of that process, an evaluation was made of the potential hazards with the new project compared to the existing and historical work and associated hazards in the affected facility

  12. Construction of a Post-Irradiated Fuel Examination Shielded Enclosure Facility

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Lehto, Ph.D.; Boyd D. Christensen

    2008-05-01

    construction activities, which will include facility modifications and construction of one shielded enclosure. Follow-up activities will be to construct two additional shielded enclosures to complete the suite of three separate but connected remote operated examination areas. Equipment purchases are to be capital procurement spread out over several years on a funded schedule. This paper discusses safety and operational considerations given during the conceptual design phase of the project. The paper considers such things as project material at risk (MAR), new processes and equipment, potential hazards, and the major modification evaluation process to determine if a preliminary Documented Safety Analysis (PDSA) is required. As part of that process, an evaluation was made of the potential hazards with the new project compared to the existing and historical work and associated hazards in the affected facility.

  13. Au Foil Activation Measurement and Simulation of the Concrete Neutron Shielding Ability for the Proposed New SANRAD Facility

    Science.gov (United States)

    Radebe, M. J.; Korochinsky, S.; Strydom, W. J.; De Beer, F. C.

    The purpose of this study was to measure the effective neutron shielding characteristics of the new shielding material designed and manufactured to be used for the construction of the new SANRAD facility at Necsa, South Africa, through Au foil activation as well as MCNP simulations. The shielding capability of the high density shielding material was investigated in the worst case region (the neutron beam axis) of the experimental chamber for two operational modes. The everyday operational mode includes the 15 cm thick poly crystalline Bismuth filter at room temperature (assumed) to filter gamma-rays and some neutron spectrum energies. The second mode, dynamic imaging, will be conducted without the Bi-filter. The objective was achieved through a foil activation measurement at the current SANRAD facility and MCNP calculations. Several Au foilswere imbedded at different thicknesses(two at each position) of shielding material up to 80 cm thick to track the attenuation of the neutron beam over distance within the shielding material. The neutron flux and subsequently the associated dose rates were calculated from the activation levels of the Au foils. The concrete shielding material was found to provide adequate shielding for all energies of neutrons emerging from beam port no-2 of the SAFARI-1 research reactorwithin a thickness of 40 cm of concrete.

  14. Beam line shielding calculations for an Electron Accelerator Mo-99 production facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-03

    The purpose of this study is to evaluate the photon and neutron fields in and around the latest beam line design for the Mo-99 production facility. The radiation dose to the beam line components (quadrupoles, dipoles, beam stops and the linear accelerator) are calculated in the present report. The beam line design assumes placement of two cameras: infra red (IR) and optical transition radiation (OTR) for continuous monitoring of the beam spot on target during irradiation. The cameras will be placed off the beam axis offset in vertical direction. We explored typical shielding arrangements for the cameras and report the resulting neutron and photon dose fields.

  15. Dose estimation and shielding calculation for X-ray hazard at high intensity laser facilities

    Science.gov (United States)

    Qiu, Rui; Zhang, Hui; Yang, Bo; James, C. Liu; Sayed, H. Rokni; Michael, B. Woods; Li, Jun-Li

    2014-12-01

    An ionizing radiation hazard produced from the interaction between high intensity lasers and solid targets has been observed. Laser-plasma interactions create “hot” electrons, which generate bremsstrahlung X-rays when they interact with ions in the target. However, up to now only limited studies have been conducted on this laser-induced radiological protection issue. In this paper, the physical process and characteristics of the interaction between high intensity lasers and solid targets are analyzed. The parameters of the radiation sources are discussed, including the energy conversion efficiency from laser to hot electrons, hot electron energy spectrum and electron temperature, and the bremsstrahlung X-ray energy spectrum produced by hot electrons. Based on this information, the X-ray dose generated with high-Z targets for laser intensities between 1014 and 1020 W/cm2 is estimated. The shielding effects of common shielding items such as the glass view port, aluminum chamber wall and concrete wall are also studied using the FLUKA Monte Carlo code. This study provides a reference for the dose estimation and the shielding design of high intensity laser facilities.

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

    International Nuclear Information System (INIS)

    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 radiotherapy x-ray facilities

    International Nuclear Information System (INIS)

    Purpose: This paper presents the results of a series of calculations to determine buildup factors for ordinary concrete, baryte concrete, lead, steel, and iron in broad beam geometry for photons energies from 0.125 to 25.125 MeV at 0.250 MeV intervals.Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials.Results: The computation of the primary broad beams using buildup factors data was done for nine published megavoltage photon beam spectra ranging from 4 to 25 MV in nominal energies, representing linacs made by the three major manufacturers. The first tenth value layer and the equilibrium tenth value layer are calculated from the broad beam transmission for these nine primary megavoltage photon beam spectra.Conclusions: The results, compared with published data, show the ability of these buildup factor data to predict shielding transmission curves for the primary radiation beam. Therefore, the buildup factor data can be combined with primary, scatter, and leakage x-ray spectra to perform computation of broad beam transmission for barriers in radiotherapy shielding x-ray facilities

  18. Analysis of shielding calculation methods for 16- and 64-slice computed tomography facilities

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, C; Cenizo, E; Bodineau, C; Mateo, B; Ortega, E M, E-mail: c_morenosaiz@yahoo.e [Servicio de RadiofIsica Hospitalaria, Hospital Regional Universitario Carlos Haya, Malaga (Spain)

    2010-09-15

    The new multislice computed tomography (CT) machines require some new methods of shielding calculation, which need to be analysed. NCRP Report No. 147 proposes three shielding calculation methods based on the following dosimetric parameters: weighted CT dose index for the peripheral axis (CTDI{sub w,per}), dose-length product (DLP) and isodose maps. A survey of these three methods has been carried out. For this analysis, we have used measured values of the dosimetric quantities involved and also those provided by the manufacturer, making a comparison between the results obtained. The barrier thicknesses when setting up two different multislice CT instruments, a Philips Brilliance 16 or a Philips Brilliance 64, in the same room, are also compared. Shielding calculation from isodose maps provides more reliable results than the other two methods, since it is the only method that takes the actual scattered radiation distribution into account. It is concluded therefore that the most suitable method for calculating the barrier thicknesses of the CT facility is the one based on isodose maps. This study also shows that for different multislice CT machines the barrier thicknesses do not necessarily become bigger as the number of slices increases, because of the great dependence on technique used in CT protocols for different anatomical regions.

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

    International Nuclear Information System (INIS)

    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

  20. Establishment of scatter factors for use in shielding calculations and risk assessment for computed tomography facilities

    International Nuclear Information System (INIS)

    The specification of shielding for CT facilities in the UK and many other countries has been based on isodose scatter curves supplied by the manufacturers combined with the scanner's mAs workload. Shielding calculations for radiography and fluoroscopy are linked to a dose measurement of radiation incident on the patient called the kerma–area product (KAP), and a related quantity, the dose-length product (DLP), is now employed for assessment of CT patient doses. In this study the link between scatter air kerma and DLP has been investigated for CT scanners from different manufacturers. Scatter air kerma values have been measured and scatter factors established that can be used to estimate air kerma levels within CT scanning rooms. Factors recommended to derive the scatter air kerma at 1 m from the isocentre are 0.36 µGy (mGy cm)−1 for the body and 0.14 µGy (mGy cm)−1 for head scans. The CT scanner gantries only transmit 10% of the scatter air kerma level and this can also be taken into account when designing protection. The factors can be used to predict scatter air kerma levels within a scanner room that might be used in risk assessments relating to personnel whose presence may be required during CT fluoroscopy procedures.

  1. 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)

  2. A decade of radiological and shielding experience at the fast flux test facility

    International Nuclear Information System (INIS)

    This paper reports on which the Fast Flux Test Facility (FFTF) which has operated for almost a decade after first going critical during February 1980. Based on about 2,000 effective full-power days of operation, it is concluded that radiological safety can be achieved in large liquid metal-cooled fast reactors. The collective dose equivalents received by operating personnel are significantly lower than those received at commercial light water reactors. No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactive materials to the environment has been well below acceptable limits. All shields have performed satisfactorily and in agreement with design calculations. The experience derived from the design, construction, and operation of the FFTF should be of inestimable value in supporting future development of liquid metal reactors

  3. Facile preparation of lightweight microcellular polyetherimide/graphene composite foams for electromagnetic interference shielding.

    Science.gov (United States)

    Ling, Jianqiang; Zhai, Wentao; Feng, Weiwei; Shen, Bin; Zhang, Jianfeng; Zheng, Wen ge

    2013-04-10

    We report a facile approach to produce lightweight microcellular polyetherimide (PEI)/graphene nanocomposite foams with a density of about 0.3 g/cm3 by a phase separation process. It was observed that the strong extensional flow generated during cell growth induced the enrichment and orientation of graphene on cell walls. This action decreased the electrical conductivity percolation from 0.21 vol % for PEI/graphene nanocomposite to 0.18 vol % for PEI/graphene foam. Furthermore, the foaming process significantly increased the specific electromagnetic interference (EMI) shielding effectiveness from 17 to 44 dB/(g/cm3). In addition, PEI/graphene nanocomposite foams possessed low thermal conductivity of 0.065-0.037 W/m·K even at 200 °C and high Young's modulus of 180-290 MPa. PMID:23465462

  4. Radiation shielding test for hot cells of Irradiated Material Examination Facility(IMEF)

    International Nuclear Information System (INIS)

    Radiation shielding test for IMEF(Irradiated Material Examination Facility) hot cell walls was executed using two Co 60 sources with the activities of 1,600 Ci and 30 Ci respectively. The tested walls are made of heavy concrete or lead, with the maximum thickness of 1,200 mm for concrete cell and 200 mm for lead cell. At first, we measured the dose rates for several standard walls and the result was used as standard reference. We also measured dose rates for hot cell walls by the same method and compared with reference. The number of testing points are 6,000 and we found out defect for several points which are mostly located in boundaries between embedded material and concrete. The defective areas were re tested after repaired and results for the areas were acceptable

  5. Practical evaluation of the biological shielding effectiveness of the Gamma Irradiation Facility at Kwabenya, Ghana

    International Nuclear Information System (INIS)

    The ability of re-inforced concrete to attenuate photons in a facility housing a 50 kCi Cobalt-60 source has been practically assessed using a Chinese made dose rate survey meter model SG-102. Just within the maze entrances, the measured dose rates were 6273.8 ± 745.7 nGy/hr (Personnel) and 755.4 ± 94.4nGy/hrs(Goods).Outside the maze entrances and behind the lead shielding doors, the net dose rates were 392.3 nGy/hr (Personnel) and 388.4 nGy/hr (Goods). At other locations within the facility, the net dose rates determined were as follows: 303.4 nGy/hr (concrete roof top), 3984 nGy/hr over the small water pool in the de-ionizer room, less than 11 nGy/hr in the control room, less than 10 nGy/hr in the electrical room and less than 3 nGy/hr in the de-ionizer room. These measured values are consistent with earlier theoretical estimates undertaken by Emi-Reynolds and Akaho. (author). 6 refs. 4 figs., 5 tabs

  6. Development of a novel interim bulk fuel storage facility for the PBMR / W.F. Fuls

    OpenAIRE

    Fuls, Wilhelm Franz

    2004-01-01

    The PBMR is the first High Temperature Reactor being designed for commercial power generation in South Africa. It makes use of spherical fuel elements, containing coated uranium oxide particles encapsulated in a graphite matrix. The spent fuel generated from the reactor is stored in a storage system before final disposal. Such storage systems are called interim storage facilities, and normally make use of small transportable containers. The PBMR design makes use of bulk storage containers...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. Numerical benchmarks TRIPOLI - MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

    International Nuclear Information System (INIS)

    3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI - MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI - MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

  10. Numerical benchmarks TRIPOLI - MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Lee, Yi-Kang [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Zeng Qin; Zhang Junjun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Serikov, Arkady [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (Germany); Trama, Jean-Christophe; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI - MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI - MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

  11. Effect of heat cycling on microstructure and thermal property of boron carbide sintered bulk as a shielding material for fusion blanket

    International Nuclear Information System (INIS)

    In the Force Free Helical Reactor (FFHR) design activity in NIFS, metallic carbides and hydrides are considered as candidate shielding materials for the fusion blankets. These materials are expected to have some advantages on neutronic and thermo-physical properties. In order to promote the blanket design, it is necessary to clarify thermal properties of the candidate materials. We studied microstructure and thermal property of boron carbide (B4C), which is one of the promising candidates shielding materials, including the effect of heat cycling. By the laser-flash method, thermal diffusivity, which is one of the properties necessary for evaluating thermal conductivity, was measured precisely for B4C samples. The thermal diffusivity of B4C around 200degC decreased to 1/3 (5 × 10-6 m2 S-1) compared with that at room temperature. The sintering density of B4C bulk was decreased slightly by the thermal cycling. It was suggested that the B4C bulk has high thermal stability and soundness of microstructure during the life-time of blanket system. (author)

  12. 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)

    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 3.7x107 MBq of 137Cs and its contribution in operating gallery was estimated

  13. Neutron shielding analysis for remote handled transuranic waste containers in facility casks at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Neutron shielding characteristics of the Waste Isolation Pilot Plant facility cask have been quantified for a variety of combinations of neutron sources and waste matrices which would potentially be handled in waste containers. The neutron attenuation and neutron environment of the waste container and the facility cask have been analyzed to ensure that the design requirement of neutron dose rate will be met under the combinations of the source and waste matrix conditions. The analyses considered the ranges of neutron source spectrum and waste matrices which combine to produce the minimum neutron shielding worth of the facility cask. One-dimensional analyses were performed with discrete ordinate transport theory methods using multigroup neutron cross section data. The results discussed in this report demonstrate the effect of source spectrum and waste container matrix on predicted neutron dose rates adjacent to the unshielded waste container and the surface of the facility cask. An evaluation of the uncertainties in predicted neutron dose rates is provided which results in an assessment of the maximum measured neutron dose rate external to the facility cask. A description of the analytical models developed, the analysis methodology, the neutron source spectra, and the detailed results are described in this report. 10 refs., 50 figs., 39 tabs

  14. SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S; Shin, E H; Kim, J; Ahn, S H; Chung, K; Kim, D-H; Han, Y; Choi, D H [Samsung Medical Center, Seoul (Korea, Republic of)

    2015-06-15

    Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using the production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  17. Shielding material

    International Nuclear Information System (INIS)

    The present invention effectively utilizes iron reinforced concrete wastes generated upon dismantling of concretes of nuclear facilities, to provide shielding material. That is, at least one of members selected from the group consisting of iron rods in iron-reinforced concretes and, regenerated aggregates regenerated from concrete wastes upon dismantling is charged in a predetermined mold. Cement pastes or cement mortars are charged therein, and solidified, cured and released from the mold. With such procedures, a block-formed shielding materials made of precast concretes can be obtained. In this case, the cements including much water of crystallization are used. Since iron reinforcing dusts and iron reinforcing dust chips are contained in the shielding materials, a great γ-ray shielding effect can be obtained. Further, since cements containing a great amount of water of crystallization are used, a great neutron shielding effect can be obtained. (I.S.)

  18. Assessment of radiation shield integrity of DD/DT fusion neutron generator facilities by Monte Carlo and experimental methods

    Science.gov (United States)

    Srinivasan, P.; Priya, S.; Patel, Tarun; Gopalakrishnan, R. K.; Sharma, D. N.

    2015-01-01

    DD/DT fusion neutron generators are used as sources of 2.5 MeV/14.1 MeV neutrons in experimental laboratories for various applications. Detailed knowledge of the radiation dose rates around the neutron generators are essential for ensuring radiological protection of the personnel involved with the operation. This work describes the experimental and Monte Carlo studies carried out in the Purnima Neutron Generator facility of the Bhabha Atomic Research Center (BARC), Mumbai. Verification and validation of the shielding adequacy was carried out by measuring the neutron and gamma dose-rates at various locations inside and outside the neutron generator hall during different operational conditions both for 2.5-MeV and 14.1-MeV neutrons and comparing with theoretical simulations. The calculated and experimental dose rates were found to agree with a maximum deviation of 20% at certain locations. This study has served in benchmarking the Monte Carlo simulation methods adopted for shield design of such facilities. This has also helped in augmenting the existing shield thickness to reduce the neutron and associated gamma dose rates for radiological protection of personnel during operation of the generators at higher source neutron yields up to 1 × 1010 n/s.

  19. Assessment of radiation shield integrity of DD/DT fusion neutron generator facilities by Monte Carlo and experimental methods

    International Nuclear Information System (INIS)

    DD/DT fusion neutron generators are used as sources of 2.5 MeV/14.1 MeV neutrons in experimental laboratories for various applications. Detailed knowledge of the radiation dose rates around the neutron generators are essential for ensuring radiological protection of the personnel involved with the operation. This work describes the experimental and Monte Carlo studies carried out in the Purnima Neutron Generator facility of the Bhabha Atomic Research Center (BARC), Mumbai. Verification and validation of the shielding adequacy was carried out by measuring the neutron and gamma dose-rates at various locations inside and outside the neutron generator hall during different operational conditions both for 2.5-MeV and 14.1-MeV neutrons and comparing with theoretical simulations. The calculated and experimental dose rates were found to agree with a maximum deviation of 20% at certain locations. This study has served in benchmarking the Monte Carlo simulation methods adopted for shield design of such facilities. This has also helped in augmenting the existing shield thickness to reduce the neutron and associated gamma dose rates for radiological protection of personnel during operation of the generators at higher source neutron yields up to 1 × 1010 n/s

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

    International Nuclear Information System (INIS)

    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

  1. 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)

    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.)

  2. Using laser entrance hole shields to increase coupling efficiency in indirect drive ignition targets for the National Ignition Facility

    International Nuclear Information System (INIS)

    Coupling efficiency, the ratio of the capsule absorbed energy to the driver energy, is a key parameter in ignition target designs. The hohlraum originally proposed for the National Ignition Facility (NIF) [G. H. Miller, E. I. Moses, and C. R. Wuest, Nucl. Fusion 44, S228 (2004)] coupled ∼11% of the absorbed laser energy to the capsule as x rays. Described here is a second generation of the hohlraum target which has a higher coupling efficiency, ∼16%. Because the ignition capsule's ability to withstand three-dimensional effects increases rapidly with absorbed energy, the additional energy can significantly increase the likelihood of ignition. The new target includes laser entrance hole (LEH) shields as a principal method for increasing coupling efficiency while controlling symmetry in indirect-drive inertial confinement fusion. The LEH shields are high Z disks placed inside the hohlraum on the symmetry axis to block the capsule's view of the relatively cold LEHs. The LEH shields can reduce the amount of laser energy required to drive a target to a given temperature via two mechanisms: (1) keeping the temperature high near the capsule pole by putting a barrier between the capsule and the pole; (2) because the capsule pole does not have a view of the cold LEHs, good symmetry requires a shorter hohlraum with less wall area. Current integrated simulations of this class of target couple 140 kJ of x rays to a capsule out of 865 kJ of absorbed laser energy and produce 10 MJ of yield. In the current designs, which continue to be optimized, the addition of the LEH shields saves ∼95 kJ of energy (about 10%) over hohlraums without LEH shields

  3. Facile synthesis of tin phosphite nanosheets via exfoliated bulk crystals: Electronic structure and piezoelectric property.

    Science.gov (United States)

    Song, Jun-Ling; Zhang, Xi-Rui; Lu, Rui-Feng

    2016-08-01

    Tin phosphite nanosheets were synthesized by a facile exfoliation method. SnHPO3 nanosheets with a thickness of ∼2.6nm readily form a stable colloidal suspension in ethanol using ultrasonic method. Structures and optical properties of the obtained nanosheets were investigated. The prepared SnHPO3 nanosheets exhibit an obvious blue-shift in UV absorbance compared with bulk SnHPO3 crystal materials. Moreover, the piezoelectric coefficients of SnHPO3 monolayer were calculated based on density functional theory, which are larger than that of h-BN monolayer, indicating this material could be a good candidate for designing electro-optical nano-devices. PMID:27175829

  4. Physical characteristics of germanium to be considered in the radiation shielding calculation for the prompt gamma rays facility analysis

    International Nuclear Information System (INIS)

    Physical characteristics of germanium (gamma detector material) to be considered in the shielding design have been researched for the prompt gammas rays facility. Reports on gamma spectrum analysis produced by neutronic capture show that an important factor to be considered are those produced by inelastic collisions with the detector material, the germanium. These collisions tend to damage the crystal and consequently, the resolution and life of detector. A research has been done concerning which energies should be considered using graphics of efficient sections existing at internet and with the MCNP calculation code

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

    International Nuclear Information System (INIS)

    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).

  6. 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)

  7. Facile synthesis of Ag-reduced graphene oxide hybrids and their application in electromagnetic interference shielding

    Science.gov (United States)

    Long, Tao; Hu, Li; Dai, HongXia; Tang, YuXia

    2014-07-01

    A fast and environmentally friendly method was proposed toward one-pot synthesis of Ag-reduced graphene oxide (Ag-RGO) hybrids by a chemical reduction method assisted by microwave irradiation treatment with the use of sodium citrate as green reductant. The as-synthesized samples were characterized systematically, and the results indicated the successful synthesis of Ag-RGO. Ag-RGO was further applied as filler in polymethyl methacrylate (PMMA) matrix polymer composites, and their electromagnetic interference (EMI) shielding performance was investigated. The prepared Ag-RGO/PMMA composites with 3.0 vol% Ag-RGO exhibited an excellent EMI shielding effectiveness (EMI SE) of average 26.8 dB in the 8-12 GHz X-band range, which outperformed the RGO/PMMA composites (18.4 dB) with bare RGO as fillers.

  8. Conservative method for determination of material thickness used in shielding of veterinary facilities; Metodo conservativo para determinacao de espessura de materiais utilizados para blindagem de instalacoes veterinarias

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    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.

  9. Experimental investigation on radiation shielding of high performance concrete for nuclear and radiotherapy facilities

    Science.gov (United States)

    Domański, Szymon; Gryziński, Michał A.; Maciak, Maciej; Murawski, Łukasz; Tulik, Piotr; Tymińska, Katarzyna

    2016-06-01

    This paper presents the set of procedures developed in Radiation Protection Measurements Laboratory at National Centre for Nuclear Research for evaluation of shielding properties of high performance concrete. The purpose of such procedure is to characterize the material behaviour against gamma and neutron radiation. The range of the densities of the concrete specimens was from 2300 to 3900 kg/m3. The shielding properties against photons were evaluated using 137Cs and 60Co sources. The neutron radiation measurements have been performed by measuring the transmitted radiation from 239PuBe source. Scattered neutron radiation has been evaluated using the shadow cone technique. A set up of ionization chambers was used during all experiments. The gamma dose was measured using C-CO2 ionization chamber. The neutron dose was evaluated with recombination chamber of REM-2 type with appropriate recombination method applied. The method to distinguish gamma and neutron absorbed dose components in mixed radiation fields using twin detector method was presented. Also, recombination microdosimetric method was applied for the obtained results. Procedures to establish consecutive half value layers and tenth value layers (HVL and TVL) for gamma and neutron radiation were presented. Measured HVL and TVL values were linked with concrete density to highlight well known dependence. Also, influence of specific admixtures to concrete on neutron attenuation properties was studied. The results confirmed the feasibility of approach for the radiation shielding investigations.

  10. Empirical shielding design data for facilities administering I131 for thyroid carcinoma

    International Nuclear Information System (INIS)

    Retrospective review of the records for 434 post thyroidectomy patients receiving I131 therapy for thyroid carcinoma revealed approximately 75% of the patients were discharged within 48 hours and 90% within 72 hours. Criterion for discharge was an external radiation dose below 25 μSv/hr, measured at one metre anterior to the patient's neck. The time-averaged average dose rate at one metre anterior to the neck of a typical patient during the isolation period was 72 μSv/hr, with 90% of the patients below 82 μSv/hr. After correcting for the effects of patient size and scatter, the effective design dose rate from a patient in an isolation room treating two or three patients/week is 105 μSv.m2.hr-1, or 75 μSv.m2.hr-1 where only one patient is treated each week. Concrete is the most economical shielding material, with 190 mm filled concrete block walls and 150+mm concrete floors as the minimum recommended shielding for a radioiodine therapy suite. Additional shielding will be required if the suite adjoins (including areas immediately above and below) areas with a high occupancy factor. Copyright (1998) Australasian Physical and Engineering Sciences in Medicine

  11. Pursuit of improvement in uranium bulk analysis at the clear facility for safeguards environmental samples

    International Nuclear Information System (INIS)

    Full text: In order to contribute to the IAEA strengthened safeguards system, a project started in Japan Atomic Energy Research Institute (JAERI) in 1998. Consequently, a clean room facility called as CLEAR, the Clean Laboratory for Environmental Analysis and Research, was constructed in June 2001 at JAERI Tokai and the analytical techniques of ultra-trace nuclear materials in environmental samples are being developed. As for the bulk analysis, performance of inductively-coupled plasma mass spectrometry (ICP-MS) was mainly examined because sample preparation for ICP-MS is simpler than that for thermal ionization mass spectrometry (TIMS). Interference of polyatomic ion (such as PtAr+) and coexisting element (such as Na) on the uranium ions, as well as mass bias caused by ICP-MS operating conditions, has been investigated for precise measurement on uranium isotope ratio. The authors have also studied on the uranium blanks during sample treatment process. The blank value below 10 pg uranium per sample treatment was obtained: dominant origins were elution from Teflon vessel surface in acid heating process of the sample to dry up. The work is in progress to minimize the blank. Compared with the process blank and the minimum uranium amount for isotope ratio measurement by ICP-MS (ca. 10 pg for natural uranium), the swipe cotton (Texwipe-304) which is currently used for IAEA Environmental Sampling includes much more amount of natural uranium in several nano-grams. If the amount of uranium collected on Texwipe-304 is small, sensitive and reliable measurement on isotope ratio will be impossible by bulk analysis. The authors are seeking alternative swipe materials with less amount of uranium. Recently, one of the authors devised an effective technique for recovery of uranium-containing particles from Texwipe-304. The technique, named as Vacuum Suction Method, uses a combination of polycarbonate membrane filters and a macro-pipette tip, which is connected to a vacuum pump

  12. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 108 n/ cm2/ s. According to IAEA (2001) flux of 1.00 x 109 n/ cm2/ s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  13. Evaluation of Brachytherapy Facility Shielding Status in Korea Obtained From Radiation Safety Reports

    International Nuclear Information System (INIS)

    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

  14. 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 ostatní: 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

  15. Utilization of radiation facilities at TNRC for shielding researches and related topics

    International Nuclear Information System (INIS)

    This paper presents the running shielding research activities at Tajura Nuclear research center. The main area of researches are concentrated on the investigation of different types of concrete made from local materials such as conventional concrete, Magnetite-Limonite concrete, and heat resistant concrete. The measuring techniques used were neutron-gamma spectrometry, and activation foils. The measurements were performed using collimated beam of reactor neutrons emitted from one of the horizontal channels, as well as from californium-252 neutron source. The transmitted neutron spectra through concrete barriers of different thicknesses were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. A non-destructive testing of some reactor materials were also carried out using neutron and gamma ray computerized tomography technique (CT). Some experiments were also carried out related to measurements of neutron depth dose distributions inside tissue equivalent materials. 10 figs

  16. Calculation method of radiation shielding in the nuclear medicine facility. Evaluation based on the reasonable calculation method

    International Nuclear Information System (INIS)

    According to the acceptance of ICRP Publication 60 (1990), the dose equivalent limit for the boarder of controlled area will be defined as 1.3 mSv/3 months in the Regulation for the Enforcement of the Medical Service Law which is scheduled to be revised. The calculating methods of radiation shielding to be considered are as follows: The first method is calculating the dose equivalent for each nuclide using 3-month maximum estimated use dose. The second method is calculating the dose equivalent using 3-month maximum estimated use dose after the conversion of all nuclide dose into that of 131I. The third method is calculating the dose equivalent using 1 day maximum estimated use dose after the conversion of all nuclide dose into that of 131I. We've investigated which of methods can meet the new regulation value (1.3 mSv/3 months). In modeled facility, we've tried to calculate the dose by the first method to confirm if we can perform the reasonable control in safe. Total dose equivalent for the boarder of controlled area (B) was 883 μSv/3 months by the first method, and its value turned out to be about 1/4 of that of the third method. Only the result by the first method was found to be within the confines of new dose equivalent limit of 1.3 mSv/3 months. The results of both method the second and the third were found to be within the confines of existing dose equivalent limit. The method as to calculate the shielding for each nuclide by using 3-month maximum estimated use dose has been accepted in the Law Concerning Prevention from Radiation Hazards due to Radioisotopes, etc. As the method is practically in accordance with the current use of radioisotope in nuclear medicine facility, the possibility of it coping with the new dose equivalent limit was indicated. (author)

  17. THE MECHANICAL AND SHIELDING DESIGN OF A PORTABLE SPECTROMETER AND BEAM DUMP ASSEMBLY AT BNLS ACCELERATOR TEST FACILITY

    International Nuclear Information System (INIS)

    A portable assembly containing a vertical-bend dipole magnet has been designed and installed immediately down-beam of the Compton electron-laser interaction chamber on beamline 1 of the Accelerator Test Facility (ATF) at Brookhaven National Laboratory (BNL). The water-cooled magnet designed with field strength of up to 0.7 Tesla will be used as a spectrometer in the Thompson scattering and vacuum acceleration experiments, where field-dependent electron scattering, beam focusing and energy spread will be analyzed. This magnet will deflect the ATF's 60 MeV electron-beam 90o downward, as a vertical beam dump for the Compton scattering experiment. The dipole magnet assembly is portable, and can be relocated to other beamlines at the ATF or other accelerator facilities to be used as a spectrometer or a beam dump. The mechanical and shielding calculations are presented in this paper. The structural rigidity and stability of the assembly were studied. A square lead shield surrounding the assembly's Faraday Cup was designed to attenuate the radiation emerging from the 1 inch-copper beam stop. All photons produced were assumed to be sufficiently energetic to generate photoneutrons. A safety evaluation of groundwater tritium contamination due to the thermal neutron capturing by the deuterium in water was performed, using updated Monte Carlo neutron-photon coupled transport code (MCNP). High-energy neutron spallation, which is a potential source to directly generate radioactive tritium and sodium-22 in soil, was conservatively assessed in verifying personal and environmental safety

  18. 40 CFR 63.11088 - What requirements must I meet for gasoline loading racks if my facility is a bulk gasoline...

    Science.gov (United States)

    2010-07-01

    ... gasoline loading racks if my facility is a bulk gasoline terminal, pipeline breakout station, or pipeline... CATEGORIES (CONTINUED) National Emission Standards for Hazardous Air Pollutants for Source Category: Gasoline... § 63.11088 What requirements must I meet for gasoline loading racks if my facility is a bulk...

  19. Shielding door

    International Nuclear Information System (INIS)

    An exhaust processing device disposed at the outside of a radioactive nuclide handling chamber is connected to a shielding door as an exit/inlet for the radioactive nuclide handling chamber. An exhaust chamber is disposed in the inside of the thick shielding door having a thickness. The exhaust chamber is always evacuated by an exhaustion blower and maintained at a negative pressure. The radioactive nuclides in the radiation nuclide handling facility are shielded by an inner seal of the double seals which seal the gap between the wall body and the shielding door. Even if a trace amount of radioactive nuclides leaks from the seal at the inner side, it is shielded by an outer seal, and sucked into the exhaust chamber which is maintained at the negative pressure. Then, it is passed from a ventilation channel through a flexible tube then caught and removed by the filter of the exhaust processing device. This can reduce the capacity of the exhaustion blower to reduce the scale of the exhaust processing device. (I.N.)

  20. Discussions for the shielding materials of synchrotron radiation beamline hutches

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  2. DEVELOPMENT OF A TAMPER RESISTANT/INDICATING AEROSOL COLLECTION SYSTEM FOR ENVIRONMENTAL SAMPLING AT BULK HANDLING FACILITIES

    Energy Technology Data Exchange (ETDEWEB)

    Sexton, L.

    2012-06-06

    Environmental sampling has become a key component of International Atomic Energy Agency (IAEA) safeguards approaches since its approval for use in 1996. Environmental sampling supports the IAEA's mission of drawing conclusions concerning the absence of undeclared nuclear material or nuclear activities in a Nation State. Swipe sampling is the most commonly used method for the collection of environmental samples from bulk handling facilities. However, augmenting swipe samples with an air monitoring system, which could continuously draw samples from the environment of bulk handling facilities, could improve the possibility of the detection of undeclared activities. Continuous sampling offers the opportunity to collect airborne materials before they settle onto surfaces which can be decontaminated, taken into existing duct work, filtered by plant ventilation, or escape via alternate pathways (i.e. drains, doors). Researchers at the Savannah River National Laboratory and Oak Ridge National Laboratory have been working to further develop an aerosol collection technology that could be installed at IAEA safeguarded bulk handling facilities. The addition of this technology may reduce the number of IAEA inspector visits required to effectively collect samples. The principal sample collection device is a patented Aerosol Contaminant Extractor (ACE) which utilizes electrostatic precipitation principles to deposit particulates onto selected substrates. Recent work has focused on comparing traditional swipe sampling to samples collected via an ACE system, and incorporating tamper resistant and tamper indicating (TRI) technologies into the ACE system. Development of a TRI-ACE system would allow collection of samples at uranium/plutonium bulk handling facilities in a manner that ensures sample integrity and could be an important addition to the international nuclear safeguards inspector's toolkit. This work was supported by the Next Generation Safeguards Initiative (NGSI

  3. Radiation shielding and safety analysis for SPring-8

    International Nuclear Information System (INIS)

    The methods of shielding design and safety analysis applied to SPring-8 are summarized. SPring-8, a third generation synchrotron radiation facility, is the facility with the highest stored electron energy of 8 GeV and very low beam emittance of 5.5 nm·rad. Because of these distinguished features, a variety of radiation issues have to be taken up, requiring the latest information for analyses. In this technical report are described the calculational methods and the conditions for the following shielding matters as well as verification of the validity; a bulk shielding, synchrotron radiation beamline shielding, skyshine, streaming through ducts and mazes, induced activities in air, cooling water and targets, and incident analysis due to abnormal beam losses. (author)

  4. Radiation shielding and safety analysis for SPring-8

    Energy Technology Data Exchange (ETDEWEB)

    Asano, Yoshihiro; Sasamoto, Nobuo [Japan Atomic Energy Research Inst., Kamigori, Hyogo (Japan). Kansai Research Establishment

    1998-03-01

    The methods of shielding design and safety analysis applied to SPring-8 are summarized. SPring-8, a third generation synchrotron radiation facility, is the facility with the highest stored electron energy of 8 GeV and very low beam emittance of 5.5 nm{center_dot}rad. Because of these distinguished features, a variety of radiation issues have to be taken up, requiring the latest information for analyses. In this technical report are described the calculational methods and the conditions for the following shielding matters as well as verification of the validity; a bulk shielding, synchrotron radiation beamline shielding, skyshine, streaming through ducts and mazes, induced activities in air, cooling water and targets, and incident analysis due to abnormal beam losses. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-02-27

    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

  6. Neutron beam experiments using nuclear research reactors: honoring the retirement of professor Bernard W. Wehring -II. 7. Redesign of the University of Texas Thermal Neutron Imaging Facility Shielding

    International Nuclear Information System (INIS)

    A thermal neutron imaging facility (TNIF) was developed at the University of Texas Nuclear Engineering Teaching Laboratory from 1994 to 1998 using a 1-MW TRIGA reactor. Currently, neutron radiography is being investigated as a method to detect flaws in large carbon composite flywheels using the TNIF. Thermal neutrons have successfully been used to detect flaws in thin carbon composites (60% of the neutrons that enter the shield walls are reflected back into the experimental area. MCNP calculations indicate that the addition of a 1.25-cm Boral liner on the inner wall is sufficient to lower the external dose to acceptable levels and reduce the percentage of neutrons reflected back into the experimental area to <2%. MCNP simulations have been a valuable tool to test shielding configurations before construction. The redesigned shutter is composed of aluminum, lead, and boron carbide. MCNP simulations for the external shielding have shown that the addition of a Boral liner on the inner shield wall is sufficient to reduce external radiation exposure to acceptable levels. The Boral liner also greatly reduces the amount of neutrons reflected back into the experimental region. The implementation of the redesigned neutron shutter and external shielding should greatly enhance the TNIF capabilities and overall usability. The new neutron shutter will allow work to be performed inside the shielding cave while the reactor is at power. The improved external shielding will enable radiographs to be taken at higher flux levels, which will be beneficial when imaging thick carbon composites. The reduction of neutron scattering within the experimental area will also enhance image quality and improve the TNIF resolution. (authors)

  7. 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; Mikellides, Ioannis; Sekerak, Michael; Polk, James

    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.

  8. Conceptual design report: below-grade bulk waste disposal facility. Formerly Utilized Sites Remedial Action Program

    International Nuclear Information System (INIS)

    This report presents two conceptual designs for below-grade land disposal facilities in the Northeastern United States for wastes managed under the Formerly Utilized Sites Remedial Action Program (FUSRAP). The wastes are low specific activity radioactive wastes generated by programs of the Manhattan Engineer District/Atomic Energy Commission (MED/AEC). One design presented is for a hypothetical disposal facility for the state of New York and one for the state of New Jersey. Each design is based on the estimated volume of FUSRAP waste in each state. Since no specific sites have been identified for the disposal facilities, the geologic, hydrologic, topographic, and meteorologic conditions chosen for the conceptual design are only representative of conditions in New York and New Jersey. The principal difference in the two sites is the assumed soil permeability which requires an engineered clay liner surrounding the waste for the New York facility, but not for the New Jersey facility. The conceptual designs are intended to be conservative and were developed to be compatible with proposed 10 CFR 61 and proposed 40 CFR 192. The designs are developed in sufficient detail to verify the feasibility of the design concepts and to provide a basis for developing capital cost estimates for below-grade land disposal facilities

  9. 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)

    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)

  10. Development of environmental sample analysis technique in KAERI. Bulk analysis and establishment of clean laboratory facility (CLASS)

    International Nuclear Information System (INIS)

    The development of analytical methods for environmental samples in Korea Atomic Energy Research Institute (KAERI) is discussed. An analysis scheme for environmental samples has been established with an MCICP-MS based bulk analysis with adopting UTEVA resin for chemical separation and a particle analysis using FTTIMS and SIMS. A clean laboratory facility called CLASS (class 100∼ class 1000) was also established in order to prevent any cross contamination of the samples. The amount of U and Pu in the process blank sample prepared in the CLASS facility was estimated as 20 pg and less than 0.005 pg, respectively. The control chart of the analytical performance for the uranium standard sample of 100 ppt (NBL U030) indicated that the analytical performance of KAERI in CLASS is within 5 % of the certified values. (author)

  11. 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)

    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)

  12. Confirmation Run of the DWPF SRAT Cycle Using the Sludge-Only Flowsheet with Tank 40 Radioactive Sludge and Frit 200 in the Shielded Cells Facility

    Energy Technology Data Exchange (ETDEWEB)

    Fellinger, T.L.

    2002-08-29

    Several basic data reports have been issued concerning the recent demonstration of the Defense Waste Processing Facility (DWPF) Sludge Receipt and Adjustment Tank (SRAT) Cycle and Slurry Mix Evaporator (SME) Cycle, conducted at the Savannah River Technology Center (SRTC). The SRTC demonstration was completed using the DWPF ''Sludge-Only'' flowsheet with washed Tank 40 sludge slurry (Sludge Batch 2 or Macrobatch 3) in the Shielded Cells facility. The DWPF ''Sludge-Only'' flowsheet calls for processing radioactive sludge slurry using nitric acid, concentrated formic acid, and frit 200.

  13. Dismantlement and removal of Old Hydrofracture Facility bulk storage bins and water tank, Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The Old Hydrofracture Facility (OHF), located at Oak Ridge National Laboratory (ORNL), was constructed in 1963 to allow experimentation and operations with an integrated solid storage, mixing, and grout injection facility. During its operation, OHF blended liquid low-level waste with grout and used a hydrofracture process to pump the waste into a deep low-permeable shale formation. Since the OHF Facility was taken out of service in 1980, the four bulk storage bins located adjacent to Building 7852 had deteriorated to the point that they were a serious safety hazard. The ORNL Surveillance and Maintenance Program requested and received permission from the US Department of Energy to dismantle the bins as a maintenance action and send the free-released metal to an approved scrap metal vendor. A 25,000-gal stainless steel water tank located at the OHF site was included in the scope. A fixed-price subcontract was signed with Allied Technology Group, Inc., to remove the four bulk storage bins and water tank to a staging area where certified Health Physics personnel could survey, segregate, package, and send the radiologically clean scrap metal to an approved scrap metal vendor. All radiologically contaminated metal and metal that could not be surveyed was packaged and staged for later disposal. Permissible personnel exposure limits were not exceeded, no injuries were incurred, and no health and safety violations occurred throughout the duration of the project. Upon completion of the dismantlement, the project had generated 53,660 lb of clean scrap metal (see Appendix D). This resulted in $3,410 of revenue generated and a cost avoidance of an estimated $100,000 in waste disposal fees.

  14. Dismantlement and removal of Old Hydrofracture Facility bulk storage bins and water tank, Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The Old Hydrofracture Facility (OHF), located at Oak Ridge National Laboratory (ORNL), was constructed in 1963 to allow experimentation and operations with an integrated solid storage, mixing, and grout injection facility. During its operation, OHF blended liquid low-level waste with grout and used a hydrofracture process to pump the waste into a deep low-permeable shale formation. Since the OHF Facility was taken out of service in 1980, the four bulk storage bins located adjacent to Building 7852 had deteriorated to the point that they were a serious safety hazard. The ORNL Surveillance and Maintenance Program requested and received permission from the US Department of Energy to dismantle the bins as a maintenance action and send the free-released metal to an approved scrap metal vendor. A 25,000-gal stainless steel water tank located at the OHF site was included in the scope. A fixed-price subcontract was signed with Allied Technology Group, Inc., to remove the four bulk storage bins and water tank to a staging area where certified Health Physics personnel could survey, segregate, package, and send the radiologically clean scrap metal to an approved scrap metal vendor. All radiologically contaminated metal and metal that could not be surveyed was packaged and staged for later disposal. Permissible personnel exposure limits were not exceeded, no injuries were incurred, and no health and safety violations occurred throughout the duration of the project. Upon completion of the dismantlement, the project had generated 53,660 lb of clean scrap metal (see Appendix D). This resulted in $3,410 of revenue generated and a cost avoidance of an estimated $100,000 in waste disposal fees

  15. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Celik, Cihangir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, Michael E [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMahan, Kimberly L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Authier, Nicolas [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Jacquet, Xavier [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Rousseau, Guillaume [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Wolff, Herve [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Savanier, Laurence [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Baclet, Nathalie [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Lee, Yi-kang [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Trama, Jean-Christophe [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Masse, Veronique [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Gagnier, Emmanuel [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Naury, Sylvie [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Blanc-Tranchant, Patrick [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Hunter, Richard [Babcock International Group (United Kingdom); Kim, Soon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dulik, George Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reynolds, Kevin H. [Y-12 National Security Complex, Oak Ridge, TN (United States)

    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

  17. Comparison of four methods used in determination of secondary shielding requirements for a teletherapy facility: a case study of 137Cs room in Tanzania

    International Nuclear Information System (INIS)

    The performance of four methods often used to calculate the secondary barrier requirements is evaluated for a typical 137Cs-therapy room as a case study. The first two methods are provided by the NCRP49 and IAEA and both consider the influence of the primary, leakage and scattered radiation at a point as corrected for the workload, use and occupancy factors. A different shielding model encompasses the third method, which determines the doses as corrected for build-up effects assuming the narrow beam geometry. The fourth method is based on the calculation of the dose rates from the source activity with a relevant gamma constant. In all four methods, an appropriate transmission factor for the protective barrier in question is applied. The results show that for controlled area, the similarity in the calculated thicknesses using all four methods was nearly within 50%. For uncontrolled areas, a significant difference of magnitude up to a factor of 2.4 was found, which is mainly attributed to the non-consideration of occupancy factors in the latter two methods. Nevertheless, the non-agreement is useful to validate the specific assumptions taken for the employed shielding method. Despite being slightly high, it is concluded that the current shielding methods based on NCRP fundamentals are satisfactorily optimal in planning new therapy facilities. However for existing facilities, such as those undesigned according to the standard requirements, the combination of the four different methods with the dose rate measurements tend to offer a better cost effective shielding option. Retrospectively, additional 41-cm thick concrete is recommended for the unshielded southern barrier of the 137Cs room. Interestingly, the recommended thickness agrees to within ±5% with that estimated by using the recently recommended method by IAEA

  18. 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.

  19. Shielding practice

    International Nuclear Information System (INIS)

    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)

  20. 40 CFR 63.11087 - What requirements must I meet for gasoline storage tanks if my facility is a bulk gasoline...

    Science.gov (United States)

    2010-07-01

    ...) If your gasoline storage tank is subject to, and complies with, the control requirements of 40 CFR... gasoline storage tanks if my facility is a bulk gasoline terminal, pipeline breakout station, or pipeline... CATEGORIES (CONTINUED) National Emission Standards for Hazardous Air Pollutants for Source Category:...

  1. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment

    International Nuclear Information System (INIS)

    The purpose of this report is to document the results from laboratory testing of the bulk vitrified (BV) waste form that was conducted in support of the 2005 integrated disposal facility (IDF) performance assessment (PA). Laboratory testing provides a majority of the key input data required to assess the long-term performance of the BV waste package with the STORM code. Test data from three principal methods, as described by McGrail et al. (2000a; 2003a), are discussed in this testing report including the single-pass flow-through test (SPFT) and product consistency test (PCT). Each of these test methods focuses on different aspects of the glass corrosion process. See McGrail et al. (2000a; 2003a) for additional details regarding these test methods and their use in evaluating long-term glass performance. In addition to evaluating the long-term glass performance, this report discusses the results and methods used to provided a recommended best estimate of the soluble fraction of 99Tc that can be leached from the engineering-scale BV waste package. These laboratory tests are part of a continuum of testing that is aimed at improving the performance of the BV waste package.

  2. Electromagnetic shielding

    International Nuclear Information System (INIS)

    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

  3. Neutron and photon doses in high energy radiotherapy facilities and evaluation of shielding performance by Monte Carlo method

    International Nuclear Information System (INIS)

    Highlights: → The MCNP5 code has been used to model a radiotherapy room of a 18 MV linear accelerator. → The neutron and the secondary gamma ray dose equivalents were evaluated at various points inside the treatment room and along the the maze. → To reduce the neutron and gamma ray doses, we have also investigated the radiotherapy room shielding performance. → The use of paraffin wax containing boron carbide indicates much better shielding effects. - Abstract: Medical accelerators operating above 10 MV are a source of undesirable neutron radiations which contaminate the therapeutic photon beam. These photoneutrons can also generate secondary gamma rays which increases undesirable dose to the patient body and to personnel and general public. In this study, the Monte Carlo N-Particle MCNP5 code has been used to model the radiotherapy room of a medical linear accelerator operating at 18 MV and to calculate the neutron and the secondary gamma ray energy spectra and the dose equivalents at various points inside the treatment room and along the maze. To validate our Monte Carlo simulation we compared our results with those evaluated by the recommended analytical methods of IAEA Report No. 47, and with experimental and simulated values published in the literature. After validation, the Monte Carlo simulation has been used to evaluate the shielding performance of the radiotherapy room. The obtained results showed that the use of paraffin wax containing boron carbide, in the lining of the radiotherapy room walls, presents enough effectiveness to reduce both neutron and gamma ray doses inside the treatment room and at the maze entrance. Such evaluation cannot be performed by the analytical methods since room material and wall surface lining are not taken into consideration.

  4. Shielding Effectiveness of Laminated Shields

    OpenAIRE

    P. V. Y. Jayasree, V. S. S. N. S. Baba, B. P. Rao

    2008-01-01

    Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigatio...

  5. Measurement and Calculation of High-Energy Neutron Spectra behind Shielding at the CERF 120 GeV/c Hadron Beam Facility

    CERN Document Server

    Nakao, N; Roesler, S; Brugger, M; Hagiwara, M; Vincke, H; Khater, H; Prinz, A A; Rokni, S H; Kosako, K

    2008-01-01

    Neutron energy spectra were measured behind the lateral shield of the CERF (CERN-EU High Energy Reference Field) facility at CERN with a 120 GeV/c positive hadron beam (a mixture of mainly protons and pions) on a cylindrical copper target (7-cm diameter by 50-cm long). An NE213 organic liquid scintillator (12.7-cm diameter by 12.7-cm long) was located at various longitudinal positions behind shields of 80- and 160-cm thick concrete and 40-cm thick iron. The measurement locations cover an angular range with respect to the beam axis between 13 and 133 degrees. Neutron energy spectra in the energy range between 32 MeV and 380 MeV were obtained by unfolding the measured pulse height spectra with the detector response functions which have been verified in the neutron energy range up to 380 MeV in separate experiments. Since the source term and experimental geometry in this experiment are well characterized and simple, and results are given in the form of energy spectra, these experimental results are very useful a...

  6. 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)

    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

  7. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

    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. A 39 neutron group self-shielded cross section library for the Lotus fusion-fission test facility

    International Nuclear Information System (INIS)

    A 39 neutron group cross section library for fusion fission blanket calculations and especially for the analysis of the LOTUS experiment has been processed using the NJOY system. The library has been generated mostly using the ENDF/B-IV basic files at 296 K. All cross sections were self-shielded using the Bondarenko method. 5 background cross sections, namely 1010, 104, 102, 10 and 1 barns respectively were considered. The tabulated dilution dependent cross sections have been interpolated with the code TRANSX-CTR which is adequate for fusion applications. The fission spectrum of the fissionable material thorium has been collapsed from the fission matrices using the Bondarenko weighting scheme. The correct geometry of the LOTUS blanket and the cell specifications were correctly considered in the interpolation scheme. Some reaction cross sections for dosimetry applications have been included into the library. These base on the more recent ENDF/B-V evaluation. Transport and response edit cross sections have been coupled in the usual way to form P0 - P3 card image tables. Furthermore they have been converted into a binary file suitable to our RSYST computational system. Moreover the cross section card image tables have been reformatted and fitted into a BXSLIB binary library for the LANL-ONEDANT transport module. (Auth.)

  9. 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)

    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 137Cs source, but present activity of the source is 1.2Ci. Hence, the shielding integrity was tested with higher active source (1.7Ci 60Co). 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

  10. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    This standard shall be applied to closed shielding facilities which, together with the lead bricks according to DIN 25 407 part 1 and the functional elements according to this standard, are designed to make possible the setting-up of complete shieldings for hot cells in beta-gamma-technique (see DIN 25 407 part 3) according to modular principles. This standard is intended to facilitate the design and construction of hot cells with shielding walls made of lead as well as the interchangeability of individual constructional elements in existing shielding walls. (orig./HP)

  11. FELIX: construction and testing of a facility to study electromagnetic effects for first wall, blanket, and shield systems

    International Nuclear Information System (INIS)

    An experimental test facility for the study of electromagnetic effects in the FWBS systems of fusion reactors has been constructed over the past 1-1/2 years at Argonne National Laboratory (ANL). In a test volume of 0.76 m3 a vertical pulsed 0.5 T dipole field (B < 50 T/s) is perpendicular to a 1 T solenoid field. Power supplies of 2.75 MW and 5.5 MW and a solid state switch rated 13 kV, 13.1 kA (170 MW) control the pulsed magnetic fields. The total stored energy in the coils is 2.13 MJ. The coils are designed for a future upgrade to 4 T or the solenoid and 1 T for the dipole field (a total of 23.7 MJ). This paper describes the design and construction features of the facility. These include the power supplies, the solid state switches, winding and impregnation of large dipole saddle coils, control of the magnetic forces, computer control of FELIX and of experimental data acquisition and analysis, and an initial experimental test setup to analyze the eddy current distribution in a flat disk

  12. 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)

  13. Shielding Effectiveness of Laminated Shields

    Directory of Open Access Journals (Sweden)

    B. P. Rao

    2008-12-01

    Full Text Available Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigations by the cavity perturbation technique, is used to compute the overall reflection and transmission coefficients of single and multiple layers of the polymers. With recent advances in synthesizing stable highly conductive polymers these lightweight mechanically strong materials appear to be viable alternatives to metals for EM1 shielding.

  14. Japanese contributions to ITER shielding neutronics design

    International Nuclear Information System (INIS)

    Shielding design for superconducting magnets and personal exposure were performed in ITER nuclear design on the basis of reports presented to the 1990 winter and summer ITER specialist meetings. Inboard shield benchmark calculation, bulk inboard shielding analysis, inboard heterogeneity effect on shielding property analysis, gap streaming analysis were discussed on shielding properties for superconducting magnets. In addition to these, transport and Monte Carlo analyses in neutral beam injector duct for biological shielding were investigated with relation to the concept of cryostat. Further biological shielding were investigated in reactor room and site boundary during the maintenance when one activated module was extracted and hanged from the ceiling. As the results of these studies, ITER shielding characteristics were evaluated and problem areas and directions for future works were shown. (author)

  15. Radiological protection issues during primary filter housing replacement in a high alpha, beta-gamma shielded facility at Dounreay

    International Nuclear Information System (INIS)

    Dounreay, on the north coast of Scotland, was home to the United Kingdom Fast Breeder Reactor (F.B.R.) development programme. F.B.R. use excess, non-moderated ('fast') neutrons to convert (breed) uranium, in elements positioned at the outer edges of the reactor core, into plutonium which can then be used as fuel.Site construction began in 1955 and three reactors were built and operated; the Dounreay Materials Test Reactor (D.M.T.R.) 1958-1969, the Dounreay Fast Reactor (D.F.R.) 1959-1977 and the Prototype Fast Reactor (P.F.R.) 1974-1994. The D.F.R. was conventionally fuelled by highly enriched uranium whereas the P.F.R. used a ceramic form of plutonium oxide (PuO2) as its fuel. Dounreay was almost entirely self-sufficient in that a fuel cycle (chemical reprocessing) area was constructed complete with recovery plants, laboratories, waste storage and other support services buildings. Liquid plutonium nitrate product was sent to Sellafield, in Cumbria, to convert for future use, and the fuel elements were then fabricated at Springfields. Eventually, recovered P.F.R. plutonium was loaded back into the reactor, closing the fuel cycle. As a matter of interest for this paper, a Post Irradiation Examination (P.I.E.) facility, D2001, was built in the early 1960' s. The Plant was equipped with a suite of ten north and south side cells built to a high level of containment within which irradiated P.F.R. fuel could be remotely disassembled and examined. This work supported the continuing development of F.B.R. design and technology and the Plant has operated very successfully throughout its lifetime. A programme of improvement was implemented to enhance reliability, productivity and to modernize the facility to meet current nuclear and engineering standards. The experience of this work is detailed in this paper. (N.C.)

  16. Shielded syringe

    International Nuclear Information System (INIS)

    This patent specification relates to a partially disposable shielded syringe for injecting radioactive material into a patient. It is claimed that the technique overcomes the problems of non-standardisation of syringe size. (U.K.)

  17. Scale-4 and related modular systems for the evaluation of nuclear facilities and package design featuring criticality, shielding and transfer capabilities

    International Nuclear Information System (INIS)

    Nuclear industry, licensing and regulatory authorities need to be able to rely on good performance of computer codes and nuclear data used in calculations for design and operation of nuclear energy facilities. Given the international impact of a major nuclear accident, and the current crisis in public confidence, it is equally important that the methods, programs and data issued should be internationally accepted. The SCALE modular system has been developed and its capabilities extended during the last 15 years. The driving idea behind its development is that it should contain well established computer codes and data libraries, have an user friendly input format, combine and automate analyses requiring multiple computer codes or calculations into standard analytic sequences and to be well documented and publicly available. The fifth version called SCALE-4 has now been released through the Radiation Shielding Information Center (RSIC) to the OECD/NEA Data Bank. SCALE is now used worldwide. The NEA Data Bank alone has distributed more than one hundred copies of the different versions. The OECD/NEA Data Bank has been asked by its international management committee to hold a seminar with the purpose of exchanging information on the latest developments and experiences among code authors and users, to ensure that users have a correct understanding as to how SCALE should be used to model different problems, and to issue recommendations for further development and benchmarking

  18. Methods for calculating radiation attenuation in shields

    International Nuclear Information System (INIS)

    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

  19. Technical specifications for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-04-01

    Technical specifications are presented concerning the safety limits and limiting safety system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls.

  20. Shielding design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report first describes the basic design philosophy of radiation shields for the fusion experimental reactor (FER) which has been proposed to be the next step machine to JT-60. Next, geometrical models and calculation parameters for shielding calculations were investigated to establish the standard design calculation methods, and accuracy of the calculation was evaluated. Further, irradiation properties of in-vessel components and bulk shielding properties were summarized in the useful form for the future design works. (author)

  1. Shield calculations, optimization vs. paradigm

    International Nuclear Information System (INIS)

    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)

  2. Shielded cells transfer automation

    International Nuclear Information System (INIS)

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. To reduce radiation exposure to operators, technological advances in remote handling and automation were employed. An industrial robot and a specially designed end effector, access port, and sealing machine were used to remotely bag waste containers out of a glove box. The system is operated from a control panel outside the work area via television cameras

  3. Radiation shielding for neutron guides

    Energy Technology Data Exchange (ETDEWEB)

    Ersez, T. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)]. E-mail: tez@ansto.gov.au; Braoudakis, G. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia); Osborn, J.C. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)

    2006-11-15

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions.

  4. Radiation protection/shield design

    International Nuclear Information System (INIS)

    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.)

  5. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor; Projeto e implantacao de melhorias na blindagem biologica da instalacao para estudos em BCNT

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Gregorio Soares de

    2011-07-01

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction {sup 10}B (n, {alpha}) {sup 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)

  6. Improvement of shielding-test-facilities for the T-tails in flutter wind-tunnel experiments%T型尾翼风洞颤振实验保护装置绕流特性分析

    Institute of Scientific and Technical Information of China (English)

    史爱明; 戎亚楠; 杨永年

    2013-01-01

    采用SST两方程湍流模型,通过求解非定常Navier-Stokes (N-S)方程,对T型尾翼风洞实验流场进行了模拟,分析了保护装置对T型尾翼风洞实验流场的影响,研究了保护装置几何外形和保护装置基座后移距离对流场影响.通过对平尾气动力的分析以及对非定常流场的对比,可以得出:采用NACA系列翼型对基座进行气动整流后,基座两侧局部超声速区显著减小,局部激波减弱甚至消失,流场品质得到改善.且采用NACA0010翼型对基座修形后的结果最理想.随着保护装置基座后移距离的增加,平尾气动力均方根值和波动值先是急剧减小,达到0.85倍平均气动弦长后开始有所增大,在2.45~4.05倍平均气动弦长范围基本不再变化,稳定到单独T型尾翼模型相应系数1倍左右.此结论对T型尾翼风洞颤振实验保护装置设计具有一定的指导意义.%The unsteady flow around the T-tails flutter model with a set of shielding-test-facilities in wind tunnel experiments were solved by applying Reynolds-Averaged Navier-Stokes (RANS) equations with SST turbulence model.Two key points were studied,one was the section shapes along airflow for shielding-test-facilities; the other was the distance between the shielding-test-facilities and the T-tails wind tunnel model.The aerodynamic bending-moments and normal-forces indexes versus the calculating time-step were presented.When a series of NACA airfoils (NACA0010-airfoil,NACA0015-airfoil,NACA0021-airfoil) were selected to enwrap the base of shielding-test-facilities,the characteristics of the flow around the T-tails flutter model with the modified shielding-test-facilities were running to quasi-steady.The strength of shock wave decreasing and the vibration of shock wave fading were deduced to the whys.The improvement of NACA0010-airfoil was much better than the others.The Root-Mean-Square (RMS)and the amplitude in the first period for the aerodynamic coefficients

  7. Simulation Studies on the New Small Wheel Shielding of the ATLAS Experiment and Design and Construction of a Test Facility for Gaseous Detectors

    OpenAIRE

    Weber, Stefan

    2016-01-01

    In this thesis two main projects are presented, both aiming at the overall goal of particle detector development. In the first part of the thesis detailed shielding studies are discussed, focused on the shielding section of the planned New Small Wheel as part of the ATLAS detector upgrade. Those studies supported the discussions within the upgrade community and decisions made on the final design of the New Small Wheel. The second part of the thesis covers the design, constructi...

  8. Normalization of shielding structure quality and the method of its studying

    International Nuclear Information System (INIS)

    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

  9. Design experience: CRBRP radiation shielding

    International Nuclear Information System (INIS)

    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''

  10. 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)

    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 Li2CO3 and the Li2O 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.)

  11. Potential Causes of Significant Inventory Differences at Bulk Handling Facilities and the Importance of Inventory Difference Action Levels

    International Nuclear Information System (INIS)

    Accountancy for nuclear material can be split into two categories. Firstly, where possible, accountancy should be in terms of items that can be transferred as discrete packages and their contents fixed at the time of their creation. All items must remain accounted for at all times, and a single missing item is considered significant. Secondly, where nuclear material is unconstrained, for example in a reprocessing plant where it can change form, there is an uncertainty that relates to the amount of material present in any location. Cumulatively, these uncertainties can be summed and provide a context for any estimate of material in a process. Any apparent loss or gain between what has been physically measured within a facility during its physical inventory take and what is reported within its nuclear material accounts is known as an inventory difference. The cumulative measurement uncertainties can be used to set an action level for the inventory difference so that if an inventory difference is observed outside of such action levels, the difference is classified as significant and an investigation to find the root cause(s) is required. The purpose of this paper is to explore the potential causes of significant inventory differences and to provide a framework within which an inventory difference investigation can be carried out.

  12. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

    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

  14. The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities

    Energy Technology Data Exchange (ETDEWEB)

    Ghassoun, J., E-mail: ghassoun@ucam.ac.ma [EPRA, Department of Physics, Faculty of Sciences Semlalia, PO Box: 2390, 40000 Marrakech (Morocco); Senhou, N. [EPRA, Department of Physics, Faculty of Sciences Semlalia, PO Box: 2390, 40000 Marrakech (Morocco)

    2012-04-15

    In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses. - Highlights: Black-Right-Pointing-Pointer The Monte Carlo method has been used to model radiotherapy room of a 18 MV linear accelerator. Black-Right-Pointing-Pointer The neutron and the gamma ray dose equivalent distributions inside a liquid (TE) phantom were evaluated. Black-Right-Pointing-Pointer The radiotherapy room shielding performance has been also investigated.

  15. The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities

    International Nuclear Information System (INIS)

    In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses. - Highlights: ► The Monte Carlo method has been used to model radiotherapy room of a 18 MV linear accelerator. ► The neutron and the gamma ray dose equivalent distributions inside a liquid (TE) phantom were evaluated. ► The radiotherapy room shielding performance has been also investigated.

  16. The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities.

    Science.gov (United States)

    Ghassoun, J; Senhou, N

    2012-04-01

    In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses. PMID:22257567

  17. Shielding Benchmark Computational Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-17

    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).

  18. Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code

    CERN Document Server

    Nunomiya, T; Nakamura, T; Nakao, N

    2002-01-01

    A 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. 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 with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...

  19. Shielding research at the Hanford Site

    International Nuclear Information System (INIS)

    The original three plutonium production reactors (B, D, and F) constructed at the Hanford Site in 1943--1944 had shields consisting of alternate layers of iron and a high-density pressed-wood product called Masonite *. This design was the engineering response to the scientific request for a mixture of iron and hydrogen. The design mix was based on earlier studies using iron and water or iron and paraffin; however, these materials did not have satisfactory structural characteristics. Although the shields performed satisfactorily, the fabrication cost was high. Each piece had to be machined precisely to fit within structural webs, so as not to introduce cracks through the shield. Before 1950, two additional reactors (DR and H) were built using the same shield design. At the request of R.L. Dickeman, an experimental facility was included in the top of the DR Reactor to permit evaluation of shield materials. Concurrent with the measurement of attenuation properties of materials in this facility, a program was undertaken to investigate the structural characteristics of various high-density Portland cement concretes. This research effort continued for over a decade, and led to the use of these concretes in subsequent reactor shields at the Hanford Site and elsewhere with significant savings in construction costs. Completion of the attenuation and structural measurements on the various high-density concretes provided a database that could be used in the design of shields for new reactors. At the Hanford Site, the top shield of the C Reactor was constructed of concrete, whereas the sides were constructed of iron-Masonite. As more and more data were acquired, the later rectors, KE, KW, and NPR, had shields of various tested concretes. Using concrete in these shields materially reduced the cost of the facilities. Additionally, studies on heat damage to the masonite resulted in changes that permitted increases in production, while at the same time maintaining shield integrity

  20. Annual evaluation of routine radiological survey/monitoring frequencies for the High Ranking Facilities Deactivating Project at Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The Bethel Valley Watershed at the Oak Ridge National Laboratory (ORNL) has several Environmental Management (EM) facilities that are designated for deactivation and subsequent decontamination and decommissioning (D and D). The Surplus Facilities Program at ORNL provides surveillance and maintenance support for these facilities as deactivation objectives are completed to reduce the risks associated with radioactive material inventories, etc. The Bechtel Jacobs Company LLC Radiological Control (RADCON) Program has established requirements for radiological monitoring and surveying radiological conditions in these facilities. These requirements include an annual evaluation of routine radiation survey and monitoring frequencies. Radiological survey/monitoring frequencies were evaluated for two High Ranking Facilities Deactivation Project facilities, the Bulk Shielding Facility and Tower Shielding Facility. Considerable progress has been made toward accomplishing deactivation objectives, thus the routine radiological survey/monitoring frequencies are being reduced for 1999. This report identifies the survey/monitoring frequency adjustments and provides justification that the applicable RADCON Program requirements are also satisfied

  1. Space reactor shielding: an assessment of the technology

    International Nuclear Information System (INIS)

    Space power reactor systems require shielding to protect payload and reactor shielding components, and also maintenance and operating personnel. Shield composition, size, and shape are important design considerations, since the shield can dominate the overall weight of the system. Techniques for space reactor shield design analysis and optimization and experimental test facilities are available for design verification. With these tools, a shielding technology in support of current and future space power reactor systems can be developed. Efforts in this direction should begin with a generic shielding program to provide information on materials properties and geometric effects and should be followed by project-specific shielding programs to provide design optimization and prototype shield verification

  2. Shielding design for target trolley for a spallation neutron source in the J-PARC project

    International Nuclear Information System (INIS)

    To pull out a mercury target horizontally and to transfer it to hot cell for replacement, a target trolley will be installed in a spallation neutron source facility in the High-intensity Proton Accelerator Project (J-PARC). According to the progress of the target trolley design and the modification of building design, the shielding performance of the mercury target trolley was evaluated. Target doses are 25 μSv/h at a manipulator operation room behind a concrete wall of 1.5 m, and 0.5 μSv/h at a non-controlled area behind another concrete wall of 1.5 m, respectively. Bending mercury piping and gaps between the target trolley and surrounding liners, etc, were modeled 3-dimensionally in order to evaluate streaming effects. Radiation doses around the target trolley were evaluated using a 3-dimensional Monte Carlo calculation code NMTC/JAM, applying the above three-dimensional model. Since concrete walls could be considered to be simple bulk shields, doses for the manipulator room and the non-controlled area were calculated using 1-dimensional spherical model with a Monte Carlo code MCNPX by using neutron fluxes at the back of the target trolley as a source. By replacing concrete shield with iron shield and reduction of gap streaming effects, the target trolley radiation shield structures were determined, which could suppress radiation doses in the manipulator room and the non-controlled area below the target doses. (author)

  3. Calculation for shielding based on the new law in the nuclear medicine facilities. Calculation methods of effective dose concerning the external and internal exposures and of radioisotope concentration concerning the exhaust gas drainage

    International Nuclear Information System (INIS)

    Following the revision of the law which incorporated the ICRP 1990 Recommendation, the medical law enforcement rule and related notices are also revised and enforced from April 1, 2001. Revised points related with the nuclear medicine facilities involve the reported items (addition of the scheduled maximum amount to be used in the next 3 months), change of dose limits at the boundary of the controlled area (from 300 μSv/w to 1.3 mSv/3 m), change of density limits in air, exhausted air and drainage, change of evaluation of radioisotope density in air (from average density during 8 hr to 1 week), change of exposure dose limits in medical workers and change of calculation method of effective dose due to internal exposure. This paper concerns the calculation methods for above and their concepts in nuclear medicine facilities in Hokkaido area. Numerical data for shielding and conditions of the facilities for clinical practice including diagnostic nuclide are taken into consideration and the actual paper forms for these items are also shown. (K.H.)

  4. Sulphate resistant shielding material

    International Nuclear Information System (INIS)

    The shielding material of the present invention is provided with sulfuric acid resistance and contains bentonite put to ion exchange treatment with barium ions as an effective ingredient. When mortars and concretes are exposed to the circumstance of sulfate, the effective ingredient functions to take place reaction between intruding sulfate and the barium ions to form insoluble barium sulfate thereby reducing chemical corrosion of mortars and concretes caused by sulfate. Cement materials, water and aggregates can optionally be contained in addition to bentonite and bentonite put to ion exchange treatment. Chemical corrosion of concretes and mortars due to intrusion of the sulfate can be prevented, and it is useful as an artificial barrier, for example, in radioactive active waste processing facilities. (T.M.)

  5. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    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/m2 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

  6. Handout on shielding calculation

    International Nuclear Information System (INIS)

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  7. Investigation of shielding analysis method for fusion reactors

    International Nuclear Information System (INIS)

    An investigation has been made, at the shielding laboratory, into the status of shielding analysis method for fusion reactor based on conceptual designs of a variety of fusion power reactors and fusion experimental facilities, in cooperation with the Fusion Reactor Shielding Working Group in the Research Committee on Fast Neutron Shielding of the Atomic Energy Society of Japan. The reactors and facilities considered are CULHAM MKII(U.K), SPTR (Japan), TFTR(U.S.A.), STARFIRE(U.S.A.) and INTOR-USA(U.S.A.). (author)

  8. Magnetic shielding design analysis

    International Nuclear Information System (INIS)

    Two passive magnetic-shielding-design approaches for static external fields are reviewed. The first approach uses the shielding solutions for spheres and cylinders while the second approach requires solving Maxwell's equations. Experimental data taken at LLNL are compared with the results from these shieldings-design methods, and improvements are recommended for the second method. Design considerations are discussed here along with the importance of material gaps in the shield

  9. Enhanced Whipple Shield

    Science.gov (United States)

    Crews, Jeanne L. (Inventor); Christiansen, Eric L. (Inventor); Williamsen, Joel E. (Inventor); Robinson, Jennifer R. (Inventor); Nolen, Angela M. (Inventor)

    1997-01-01

    A hypervelocity impact (HVI) Whipple Shield and a method for shielding a wall from penetration by high velocity particle impacts where the Whipple Shield is comprised of spaced apart inner and outer metal sheets or walls with an intermediate cloth barrier arrangement comprised of ceramic cloth and high strength cloth which are interrelated by ballistic formulae.

  10. Electromagnetically shielded building

    International Nuclear Information System (INIS)

    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

  11. Electromagnetic shielding formulae

    International Nuclear Information System (INIS)

    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)

  12. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    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)

  13. Rotating shielded crane system

    Science.gov (United States)

    Commander, John C.

    1988-01-01

    A rotating, radiation shielded crane system for use in a high radiation test cell, comprises a radiation shielding wall, a cylindrical ceiling made of radiation shielding material and a rotatable crane disposed above the ceiling. The ceiling rests on an annular ledge intergrally attached to the inner surface of the shielding wall. Removable plugs in the ceiling provide access for the crane from the top of the ceiling into the test cell. A seal is provided at the interface between the inner surface of the shielding wall and the ceiling.

  14. Space reactor shield technology

    International Nuclear Information System (INIS)

    The reactor shield mass contributes a large portion (10% to 25%) to the total mass of an unmanned reactor system. Different shield materials are required to attenuate neutrons and gamma rays and still obtain a minimum mass. The shield material selection should also consider structural characteristics, physical and chemical properties, fabricability and availability. Minimum mass is achieved by using a shadow shield. Neutron capture gamma ray and heat generation are extremely important considerations. Lithium hydride was selected for the neutron shield material due to its excellent properties. It has to be canned and may be compartmentalized to reduce the probability of complete shielding effectiveness loss due to meteoroid puncture of the can. The initial shield design was based on previous SNAP shield design experience. The Monte Carlo Neutron Photon code, which includes the radiation scattering with the radiator and power conversion system, was then used to ensure that the design requirements were met. Fabrication of the shield by casting techniques is recommended to maintain shield integrity during vibration and to accommodate complex penetrations. A method for casting full-scale shields is described

  15. Shielded canister transporter equipment acceptance test operations

    International Nuclear Information System (INIS)

    The defense waste processing facility (DWPF) processes high level waste at the Savannah River Plant (SRP) by vitrifying the waste and placing it in stainless stell canisters for long term storage. The shielded canister transporter (SCT) is a diesel powered mobile rubber tired self-propelled vehicle which transports the canisters from the DWPF processing facility to the on-site waste storage building. The SCT has a system of automatic programmable logic controls (PLC) which provides operational handling control with a shielded transfer cask and associated canister positional equipment

  16. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    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.)

  17. Verification of effectiveness of borated water shield for a cyclotron type self-shielded

    International Nuclear Information System (INIS)

    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 18F-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)

  18. Bulk undercooling

    Science.gov (United States)

    Kattamis, T. Z.

    1984-01-01

    Bulk undercooling methods and procedures will first be reviewed. Measurement of various parameters which are necessary to understand the solidification mechanism during and after recalescence will be discussed. During recalescence of levitated, glass-encased large droplets (5 to 8 mm diam) high speed temperature sensing devices coupled with a rapid response oscilloscope are now being used at MIT to measure local thermal behavior in hypoeutectic and eutectic binary Ni-Sn alloys. Dendrite tip velocities were measured by various investigators using thermal sensors or high speed cinematography. The confirmation of the validity of solidification models of bulk-undercooled melts is made difficult by the fineness of the final microstructure, the ultra-rapid evolution of the solidifying system which makes measurements very awkward, and the continuous modification of the microstructure which formed during recalescence because of precipitation, remelting and rapid coarsening.

  19. Methods and procedures for shielding analyses for the SNS

    International Nuclear Information System (INIS)

    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)

  20. Resonance self-shielding near zone interfaces

    International Nuclear Information System (INIS)

    A practical methodology is developed to treat the resonance self-shielding transition near zone interfaces. Based on the narrow resonance approximation, a space- and energy-dependent self-shielding factor for a single interface system is derived from the integral transport theory. Using the Wigner rational approximation, the self-shielding factor for a fine region near a zone interface is factorized into a linear combination of individual homogeneous and heterogeneous self-shielding factors. The method has been implemented in a widely used cross-section processing code that is based on the Bondarenko f-factor method. The result of the analysis was applied to a fast reactor blanket mock-up to improve the calculations near a converter-blanket interface. Comparisons of the calculation with /sup 238/U capture experimental data measured in the Purdue Fast Breeder Blanket Facility are also discussed

  1. Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code

    International Nuclear Information System (INIS)

    A 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. 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 with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12C(n,2n)11C 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. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The

  2. Under the Rape Shield

    OpenAIRE

    Roman, Denise

    2011-01-01

    This article focuses on the Rape Shield Laws and their evolution in the United States, one of the pioneers in this field. The article also discusses constitutional and feminist critiques of present Rape Shield Laws, and ends with a comparative perspective throughout the Anglo-American legal space today. Finally, although the Rape Shield Laws can be approached from a variety of discourses, this article engages specifically with a discourse that intersects legal and feminist analyses.

  3. Accelerator shielding benchmark problems

    International Nuclear Information System (INIS)

    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)

  4. Accelerator shielding benchmark problems

    Energy Technology Data Exchange (ETDEWEB)

    Hirayama, H.; Ban, S.; Nakamura, T. [and others

    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).

  5. Analysis on the shielding ability of a hot cell to accommodate advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    A design work is conducting for the IMEF's future cell which located in the basement to use it as a demonstration facility for Advanced Spent Fuel Conditioning Process (ACP). Since the total radiation source which used in ACP is expected as approximately 10 times higher than the design criteria of IMEF, the existing concrete structure cannot meet the shielding requirements. Therefore, shielding design which reinforcing the shielding capability has carried out for the ACP hot cell to satisfy the shielding criteria for the expected maximum radioactivity of ACP. This study presents a shielding analysis results using QADS code for the reinforced shielding wall with heavy concrete, steel or lead, etc. As a results of the analysis, a shielding wall reinforcing method was proposed. Additional shielding analysis was performed for the ACP hot cell with proposed reinforced shielding design using MCNP-4C code, and the validity of radiation shielding design was evaluated

  6. Aircraft shielding experiments at general dynamics Fort Worth

    International Nuclear Information System (INIS)

    The Nuclear Aircraft Research Facility was established by Convair, Fort Worth, in 1950 under U.S. Air Force auspices to support the Aircraft Nuclear Propulsion Program in the areas of shielding and radiation effects problems affecting the airframe. The company subsequently became General Dynamics, Fort Worth. In 1954, an experimental shielding program was developed by B.P. Leonard and N.M. Schaeffer that incorporated air, ground, and structure scattering experiments with three sources: a large Co source, the gorund test reactor (GTR), and finally, the aircraft shield test reactor (ASTR). Shield penetration measurements were also planned with the GTR. Principal elements of this program are summarized in the paper

  7. Rotating shielded crane system

    International Nuclear Information System (INIS)

    A rotating, radiation-shielded crane system is described comprising: a generally cylindrical, radiation-shielding wall, the top of the wall forming a first annular ledge; a second annular ledge integrally attached to the inner surface of the shielding wall; a generally cylindrical ceiling made of radiation shielding material, the ceiling including a flange portion on the top thereof and a body portion, the flange portion associated with the second annular ledge such that the ceiling is supported thereby, the volume inside the wall and the ceiling forming a test cell; a rotatable crane disposed above the ceiling such that the crane is outside of the test cell; removable access means in the ceiling for allowing the crane to access the inside of the test cell from the top of the ceiling; means for sealing the interface between the inner surface of the shielding wall and the ceiling

  8. Fusion Engineering Device (FED) first wall/shield design

    International Nuclear Information System (INIS)

    The torus of the Fusion Engineering Device (FED) is comprised of the bulk shield and its associated spool lstructure and support system, the first wall water-cooled panel and armor systems, and the pumped limiter. The bulk shielding is provided by ten shield sectors that are installed in the spool structure in such a way as to permit extraction of the sectors through the openings between adjacent toroidal field coils with a direct radial movement. The first wall armor is installed on the inboard and top interior walls of these sectors, and the water-cooled panels are installed on the outboard interior walls and the pumped limiter in the bottom of the sectors. The overall design of the first wall and shield system is described in this paper

  9. Asphalt as biological shielding against fusion neutrons

    International Nuclear Information System (INIS)

    For fusion experiments, thick biological radiation protection shields are necessary due to the deep penetration capability of the 14 MeV neutrons. A (D,T) neutron generator with a moderate output of around 1012 n/sec requires a concrete shielding of a wall thickness of 2 meters laterally and at the top of an experimental assembly. The cost for this biological shield may exceed the cost for most of the equipment for a fusion and/or hybrid experimental installation. Particularly, in Saudi Arabia, asphalt is very cheap and available in bulk quantities. As it is rich in hydrogen and carbon, it is worthwhile to investigate its shielding potential against fusion neutron. In the present work different biological shield configurations of asphalt at the wall of the experimental cavity for a research program being undertaken in Saudi Arabia, are investigated. The experimental cavity is approximated by a sphere of 5 meters radius. The yield of the neutron generator is taken as 1012 - 14 MeV - neutron/sec

  10. 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.

  11. Inhibited Shaped Charge Launcher Testing of Spacecraft Shield Designs

    Science.gov (United States)

    Grosch, Donald J.

    1996-01-01

    This report describes a test program in which several orbital debris shield designs were impact tested using the inhibited shaped charge launcher facility at Southwest Research Institute. This facility enables researchers to study the impact of one-gram aluminum projectiles on various shielding designs at velocities above 11 km/s. A total of twenty tests were conducted on targets provided by NASA-MSFC. This report discusses in detail the shield design, the projectile parameters and the test configuration used for each test. A brief discussion of the target damage is provided, as the detailed analysis of the target response will be done by NASA-MSFC.

  12. A facile approach to fabrication of novel CeO2–TiO2 core–shell nanocomposite leads to excellent UV-shielding ability and lower catalytic activity

    International Nuclear Information System (INIS)

    This study reports the development of a fast and facile route for the synthesis of novel CeO2–TiO2 core–shell nanocomposite particles using microwave (MW) irradiation of the mixture of commercial CeO2, titanium-tetra-n-butoxide (TBOT) and aqueous ammonia. Solutions of TBOT in ethanol and ammonia were mixed with dispersed CeO2 nanoparticles in ethanol, and the mixture was rapidly MW irradiated at 70 °C for 2 min. The resulting nanocomposite particles were characterized in terms of phase, shell thickness, composition, surface charge, morphology, and chemical state of the elements by XRD, TEM, XPS, SEM, Zeta potential analyzer, XRF, and FT-IR. Conventional methods of the synthesis of CeO2–TiO2 nanocomposite require a long time, and TiO2 is rarely found as a coated material. In contrast, the MW method was able to synthesize CeO2–TiO2 core–shell nanocompsite particles within a very short time. CeO2–TiO2 nanocomposite particles were fairly unaggregated with an average titania layer thickness of 2–5 nm. The obtained nanocomposites retained the crystalline cubic phase of CeO2, and the phase of coated TiO2 was amorphous. The catalytic activities of uncoated and TiO2-coated CeO2 nanoparticles for the oxidation of organic compounds were evaluated by the degradation study of methylene blue in air atmosphere at 403 K. The enhanced UV-shielding ability and visible transparency of the nanocomposite obtained by UV visible spectroscopic measurements suggested that the core–shell material has novel characteristics for using as a sunscreen material.

  13. A facile approach to fabrication of novel CeO{sub 2}-TiO{sub 2} core-shell nanocomposite leads to excellent UV-shielding ability and lower catalytic activity

    Energy Technology Data Exchange (ETDEWEB)

    Bahadur, Newaz Mohammed, E-mail: nmbahadur@yahoo.com [Utsunomiya University, Laboratory of Powder Technology, Graduate School of Engineering, Venture Business Laboratry (Japan); Kurayama, Fumio [Utsunomiya University, Center for Optical Research and Education (Japan); Furusawa, Takeshi; Sato, Masahide [Utsunomiya University, Department of Advanced Interdisciplinary Sciences (Japan); Siddiquey, Iqbal Ahmed [Utsunomiya University, Laboratory of Powder Technology, Graduate School of Engineering, Venture Business Laboratry (Japan); Hossain, Md. Mufazzal [University of Dhaka, Department of Chemistry (Bangladesh); Suzuki, Noboru [Utsunomiya University, Laboratory of Powder Technology, Graduate School of Engineering, Venture Business Laboratry (Japan)

    2013-01-15

    This study reports the development of a fast and facile route for the synthesis of novel CeO{sub 2}-TiO{sub 2} core-shell nanocomposite particles using microwave (MW) irradiation of the mixture of commercial CeO{sub 2}, titanium-tetra-n-butoxide (TBOT) and aqueous ammonia. Solutions of TBOT in ethanol and ammonia were mixed with dispersed CeO{sub 2} nanoparticles in ethanol, and the mixture was rapidly MW irradiated at 70 Degree-Sign C for 2 min. The resulting nanocomposite particles were characterized in terms of phase, shell thickness, composition, surface charge, morphology, and chemical state of the elements by XRD, TEM, XPS, SEM, Zeta potential analyzer, XRF, and FT-IR. Conventional methods of the synthesis of CeO{sub 2}-TiO{sub 2} nanocomposite require a long time, and TiO{sub 2} is rarely found as a coated material. In contrast, the MW method was able to synthesize CeO{sub 2}-TiO{sub 2} core-shell nanocompsite particles within a very short time. CeO{sub 2}-TiO{sub 2} nanocomposite particles were fairly unaggregated with an average titania layer thickness of 2-5 nm. The obtained nanocomposites retained the crystalline cubic phase of CeO{sub 2}, and the phase of coated TiO{sub 2} was amorphous. The catalytic activities of uncoated and TiO{sub 2}-coated CeO{sub 2} nanoparticles for the oxidation of organic compounds were evaluated by the degradation study of methylene blue in air atmosphere at 403 K. The enhanced UV-shielding ability and visible transparency of the nanocomposite obtained by UV visible spectroscopic measurements suggested that the core-shell material has novel characteristics for using as a sunscreen material.

  14. Shielding superconductors with thin films

    CERN Document Server

    Posen, Sam; Catelani, Gianluigi; Liepe, Matthias U; Sethna, James P

    2015-01-01

    Determining the optimal arrangement of superconducting layers to withstand large amplitude AC magnetic fields is important for certain applications such as superconducting radiofrequency cavities. In this paper, we evaluate the shielding potential of the superconducting film/insulating film/superconductor (SIS') structure, a configuration that could provide benefits in screening large AC magnetic fields. After establishing that for high frequency magnetic fields, flux penetration must be avoided, the superheating field of the structure is calculated in the London limit both numerically and, for thin films, analytically. For intermediate film thicknesses and realistic material parameters we also solve numerically the Ginzburg-Landau equations. It is shown that a small enhancement of the superheating field is possible, on the order of a few percent, for the SIS' structure relative to a bulk superconductor of the film material, if the materials and thicknesses are chosen appropriately.

  15. Scale-PC shielding analysis sequences

    International Nuclear Information System (INIS)

    The SCALE computational system is a modular code system for analyses of nuclear fuel facility and package designs. With the release of SCALE-PC Version 4.3, the radiation shielding analysis community now has the capability to execute the SCALE shielding analysis sequences contained in the control modules SAS1, SAS2, SAS3, and SAS4 on a MS- DOS personal computer (PC). In addition, SCALE-PC includes two new sequences, QADS and ORIGEN-ARP. The capabilities of each sequence are presented, along with example applications

  16. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    International Nuclear Information System (INIS)

    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

  17. Secondary gamma-ray data for shielding calculation

    International Nuclear Information System (INIS)

    In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

  18. Radiation shielding device

    International Nuclear Information System (INIS)

    Purpose: To lower the shielding cost by providing a shielding wall having cavities and charging spherical shiedling materials in the cavities only when the shielding is required. Constitution: The structure comprises two parallel steel side plates aparting from each other to form a space therebetween and reinforcements such as H-type steels vertically provided between the side plates. The upper and the lower ends of the reinforcements are aparted from the upper and the lower edges of the side plates by a predetermined distance to form lateral passage between the top plate and the bottom plate. A guide plate having a plurality of openings is mounted on the upper ends of the reinforcements. If it is required for the structure to serve as the shield, spherical radioactive shielding materials are supplied through an injection port onto the guide plate while opening the injection port is opened and closing discharge port. The spherical radioactive shielding materials are fallen through the openings and filled in the space to thereby providing the structure with shielding performance. (Yoshino, Y.)

  19. Experimental verification of FOREV-2D simulations for the plasma shield

    International Nuclear Information System (INIS)

    Analysis of experiments in the MK-200UG facility dedicated to verify the FOREV-2D simulations of ITER core contamination with carbon vaporized during ELMs has been performed. In these experiments the carbon fibre composite (CFC) of NB31 grade have been treated with plasma heat fluxes relevant for ITER ELMs. The analysis revealed that thin layer of few hundred microns on CFC surface is damaged and its thermoconductivity effectively reduced approximately three times, but the CFC bulk has the reference thermoconductivity. Good agreement between the measured and the calculated profiles for carbon plasma electron density at various hydrogen plasma heat loads as well as the agreement between the measured and the simulated dependences of the absorbed energy density on the applied heat load provide reliable validation of the carbon plasma shields simulated with the FOREV-2D code. High carbon plasma shield densities of 1023-1024 m-3 predicted in the simulations for ELM-produced shields has been proved in these MK-200UG experiments.

  20. Shielding member for thermonuclear device

    International Nuclear Information System (INIS)

    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. Gamma-ray shielding design and performance test of WASTEF

    International Nuclear Information System (INIS)

    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 104 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 106 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)

  2. Alternate shield material feasibility

    International Nuclear Information System (INIS)

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B4C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B4C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B4C would only be 0.002%. No adverse reactor impact would occur if the B4C escaped from the B4C shields

  3. Alternate shield material feasibility

    Energy Technology Data Exchange (ETDEWEB)

    Specht, E.R.; Levitt, L.B.

    1984-04-01

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B/sub 4/C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B/sub 4/C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B/sub 4/C would only be 0.002%. No adverse reactor impact would occur if the B/sub 4/C escaped from the B/sub 4/C shields.

  4. Consolidated fuel shielding calculations

    International Nuclear Information System (INIS)

    Irradiated fuel radiation dose rate and radiation shielding requirements are calculated using a validated ISOSHLD-II model. Comparisons are made to experimental measurements. ISOSHLD-11 calculations are documented

  5. 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...

  6. Radiation shielding curtain

    International Nuclear Information System (INIS)

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

  7. Shield for a medical actinometer

    International Nuclear Information System (INIS)

    The shield is designed for an actinometer enabling a kidney clearance determination. It shields the radioactive radiation coming from the kidney-bladder region opposite the measuring head. The shield consists of two plates which can be pushed together so that the dimensions of the shield are variable. (DG)

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

    International Nuclear Information System (INIS)

    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

  9. Subsurface Shielding Source Term Specification Calculation

    International Nuclear Information System (INIS)

    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

  10. Neutron shielding material

    International Nuclear Information System (INIS)

    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)

  11. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

    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)

  12. Optimal selection for shielding materials by fuzzy linear programming

    International Nuclear Information System (INIS)

    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

  13. SHIELD II: VLA HI Spectral Line Observations

    Science.gov (United States)

    Lee, Eojin; Cannon, John M.; McNichols, Andrew; Teich, Yaron; SHIELD II Team

    2016-01-01

    The "Survey of HI in Extremely Low-mass Dwarfs II" ("SHIELD II") is a multiwavelength, legacy-class observational campaign that is facilitating the study of both internal and global evolutionary processes in low-mass dwarf galaxies discovered by the Arecibo Legacy Fast ALFA (ALFALFA) survey. We present new results from low-resolution D-configuration VLA HI spectral line observations of 6 galaxies in the SHIELD II sample. We explore the morphology and kinematics by comparing images of the HI surface densities and the intensity weighted velocity fields with optical images from SDSS and WIYN. These data allow us to localize the HI gas and to study the bulk neutral gas kinematics.Support for this work was provided by NSF grant AST-1211683 to JMC at Macalester College.

  14. Shields-1, A SmallSat Radiation Shielding Technology Demonstration

    Science.gov (United States)

    Thomsen, D. Laurence, III; Kim, Wousik; Cutler, James W.

    2015-01-01

    The NASA Langley Research Center Shields CubeSat initiative is to develop a configurable platform that would allow lower cost access to Space for materials durability experiments, and to foster a pathway for both emerging and commercial-off-the-shelf (COTS) radiation shielding technologies to gain spaceflight heritage in a relevant environment. The Shields-1 will be Langleys' first CubeSat platform to carry out this mission. Radiation shielding tests on Shields-1 are planned for the expected severe radiation environment in a geotransfer orbit (GTO), where advertised commercial rideshare opportunities and CubeSat missions exist, such as Exploration Mission 1 (EM-1). To meet this objective, atomic number (Z) graded radiation shields (Zshields) have been developed. The Z-shield properties have been estimated, using the Space Environment Information System (SPENVIS) radiation shielding computational modeling, to have 30% increased shielding effectiveness of electrons, at half the thickness of a corresponding single layer of aluminum. The Shields-1 research payload will be made with the Z-graded radiation shields of varying thicknesses to create dose-depth curves to be compared with baseline materials. Additionally, Shields-1 demonstrates an engineered Z-grade radiation shielding vault protecting the systems' electronic boards. The radiation shielding materials' performances will be characterized using total ionizing dose sensors. Completion of these experiments is expected to raise the technology readiness levels (TRLs) of the tested atomic number (Z) graded materials. The most significant contribution of the Z-shields for the SmallSat community will be that it enables cost effective shielding for small satellite systems, with significant volume constraints, while increasing the operational lifetime of ionizing radiation sensitive components. These results are anticipated to increase the development of CubeSat hardware design for increased mission lifetimes, and enable

  15. Shielding from cosmic radiation for interplanetary missions Active and passive methods

    CERN Document Server

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

    2007-01-01

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

  16. Radiation shielding bricks

    International Nuclear Information System (INIS)

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

  17. Technical Aspect of Shielded SIMS Installation in CEA Cadarache

    International Nuclear Information System (INIS)

    A shielded IMS 6f has been installed in the LECA facility, CEA Cadarache France. The nuclearization was performed by CAMECA Company, which sells the standard IMS 6f. Working on nuclear materials requires in depth modifications of the apparatus itself. Despite these modifications, the shielded SIMS has the same level of performance as the standard apparatus. The design of the modified apparatus is presented and the safety aspects are emphasised. The shielded SIMS should be allowed to handle irradiated samples at the end of 2001. (Author)

  18. Plasma shield dynamics and target erosion in disruption simulation experiments

    International Nuclear Information System (INIS)

    Dynamics of carbon plasma shields and graphite target erosion experimentally studied at the recently upgraded plasma gun facility MK-200 UG at TRINITI Troitsk are compared with numerical results from FOREV-2 which allows a 2-D modeling of hot plasma target interaction. It is shown that turbulent processes are absent in the experimental plasma shields. Anomalous lateral losses of plasma mass are not occurring and plasma shield dynamics can be described by the multidimensional radiation magnetohydrodynamics (R-MHD) equations together with a consistent solution of the multidimensional magnetic field equations with classical magnetic field diffusion coefficient

  19. Radiation shielding materials

    International Nuclear Information System (INIS)

    Purpose: To obtain putty-like shielding materials excellent in the radiation shielding and packing workability for use in penetrations of electrical wires or pipeways in a nuclear installation. Constitution: A putty-like material is prepared from 100 parts by weight of a binder comprising a grease or the like having viscosity of greater than 5000 cst or an immiscible consistency of greater than 100 (JIS K 2220 (1980) para. 5.3.4) at 25 0C and from 1200 to 4000 parts by weight of high density inorganic powder such as lead powder or lead oxide powder having a density of greater than 5 g/cm3 and such a particle size that more than 95 % thereof passes through a 145 mesh sieve. The putty-like material is adjusted such that it has 1 - 35 mm of softness (JIS A 5752) at normal temperature, more than 1 g/5 sec of injection amount and a density of greater than 4 g/cm3. In this way, non-curable radiation shielding agent with excellent X-ray or γ-ray shielding property and being capable of packed densely to void portions can be obtained. (Ikeda, J.)

  20. Shield For Flexible Pipe

    Science.gov (United States)

    Ponton, Michael K.; Williford, Clifford B.; Lagen, Nicholas T.

    1995-01-01

    Cylindrical shield designed to fit around flexible pipe to protect nearby workers from injury and equipment from damage if pipe ruptures. Designed as pressure-relief device. Absorbs impact of debris ejected radially from broken flexible pipe. Also redirects flow of pressurized fluid escaping from broken pipe onto flow path allowing for relief of pressure while minimizing potential for harm.

  1. 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.

  2. Lightweight Shield Against Space Debris

    Science.gov (United States)

    Redmon, John W., Jr.; Lawson, Bobby E.; Miller, Andre E.; Cobb, W. E.

    1992-01-01

    Report presents concept for lightweight, deployable shield protecting orbiting spacecraft against meteoroids and debris, and functions as barrier to conductive and radiative losses of heat. Shield made in four segments providing 360 degree coverage of cylindrical space-station module.

  3. Efficacy of Cosmic Ray Shields

    Science.gov (United States)

    Rhodes, Nicholas

    2015-10-01

    This research involved testing various types of shielding with a self-constructed Berkeley style cosmic ray detector, in order to evaluate the materials of each type of shielding's effectiveness at blocking cosmic rays and the cost- and size-efficiency of the shields as well. The detector was constructed, then tested for functionality and reliability. Following confirmation, the detector was then used at three different locations to observe it altitude or atmospheric conditions had any effect on the effectiveness of certain shields. Multiple types of shielding were tested with the detector, including combinations of several shields, primarily aluminum, high-iron steel, polyethylene plastic, water, lead, and a lead-alternative radiation shield utilized in radiology. These tests regarding both the base effectiveness and the overall efficiency of shields is designed to support future space exploratory missions where the risk of exposure to possibly lethal amounts of cosmic rays for crew and the damage caused to unshielded electronics are of serious concern.

  4. Hinged Shields for Machine Tools

    Science.gov (United States)

    Lallande, J. B.; Poland, W. W.; Tull, S.

    1985-01-01

    Flaps guard against flying chips, but fold away for tool setup. Clear plastic shield in position to intercept flying chips from machine tool and retracted to give operator access to workpiece. Machine shops readily make such shields for own use.

  5. Spacecraft Electrostatic Radiation Shielding

    Science.gov (United States)

    2008-01-01

    This project analyzed the feasibility of placing an electrostatic field around a spacecraft to provide a shield against radiation. The concept was originally proposed in the 1960s and tested on a spacecraft by the Soviet Union in the 1970s. Such tests and analyses showed that this concept is not only feasible but operational. The problem though is that most of this work was aimed at protection from 10- to 100-MeV radiation. We now appreciate that the real problem is 1- to 2-GeV radiation. So, the question is one of scaling, in both energy and size. Can electrostatic shielding be made to work at these high energy levels and can it protect an entire vehicle? After significant analysis and consideration, an electrostatic shield configuration was proposed. The selected architecture was a torus, charged to a high negative voltage, surrounding the vehicle, and a set of positively charged spheres. Van de Graaff generators were proposed as the mechanism to move charge from the vehicle to the torus to generate the fields necessary to protect the spacecraft. This design minimized complexity, residual charge, and structural forces and resolved several concerns raised during the internal critical review. But, it still is not clear if such a system is costeffective or feasible, even though several studies have indicated usefulness for radiation protection at energies lower than that of the galactic cosmic rays. Constructing such a system will require power supplies that can generate voltages 10 times that of the state of the art. Of more concern is the difficulty of maintaining the proper net charge on the entire structure and ensuring that its interaction with solar wind will not cause rapid discharge. Yet, if these concerns can be resolved, such a scheme may provide significant radiation shielding to future vehicles, without the excessive weight or complexity of other active shielding techniques.

  6. Internally shielded beam transport and support system

    International Nuclear Information System (INIS)

    Due to environmental concerns, the Advanced Photon Source has a policy that disallows any exposed lead within the facility. This creates a real problem for the beam transport system, not so much for the pipe but for the flexible coupling (bellows) sections. A complete internally shielded x-ray transport system, consisting of long transport lines joined by flexible coupling sections, has been designed for CARS sector 14 to operate either at high vacuum or as a helium flight tube. It can effectively shield against air scattering of wiggler or undulator white beam with proper placement of apertures, collimators, and masks for direct beam control. The system makes use of male- and female-style fittings that create a labyrinth allowing for continuous shielding through the flexible coupling sections. These parts are precision machined from a ternary hypereutectic lead alloy (cast under 15 inches of head pressure to assure a pinhole-free casting) then pressed into either end (rotatable vacuum flanges) of the bellows assembly. The transport pipe itself consists of a four part construction using a stepped transition ring (Z-ring) to connect an inner tube to the vacuum flange and also to a protective and supportive outer tube. The inner tube is wrapped with 1/16 double-prime pure lead sheet to a predetermined thickness following the shape of the stepped transition ring for continuous shielding. This design has been prototyped and radiation tested. copyright 1996 American Institute of Physics

  7. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

    Cherkashyna, Nataliia; DiJulio, Douglas D.; Panzner, Tobias; Rantsiou, Emmanouela; Filges, Uwe; Ehlers, Georg; Bentley, Phillip M.

    2015-01-01

    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...

  9. Attenuation curves in concrete of neutrons from 1 GeV/u C and U ions on a Fe target for the shielding design of RIB in-flight facilities

    CERN Document Server

    Agosteo, S; Silari, M

    2004-01-01

    Experimental data on neutron emission from the interaction of heavy ion beams with matter are far less abundant than data on neutron production from protons. The aim of the present work is to extend the available computational shielding data to high-energy neutrons produced by heavy ion beams (uranium and carbon) of 1 GeV/u slowed down in a thick iron target. Source terms and attenuation lengths for neutron attenuation in a concrete shield were calculated starting from experimental neutron energy distributions measured at GSI in the angular range from 0 degree to 90 degree . A comparison is also made with previous calculations performed for different ions and energies and with earlier estimates made at GSI for neon beams with 0.8 and 2 GeV/u energy stopped in thick iron, lead and uranium targets.

  10. Capacitive Proximity Sensors With Additional Driven Shields

    Science.gov (United States)

    Mcconnell, Robert L.

    1993-01-01

    Improved capacitive proximity sensors constructed by incorporating one or more additional driven shield(s). Sensitivity and range of sensor altered by adjusting driving signal(s) applied to shield(s). Includes sensing electrode and driven isolating shield that correspond to sensing electrode and driven shield.

  11. Multilayer radiation shield

    Science.gov (United States)

    Urbahn, John Arthur; Laskaris, Evangelos Trifon

    2009-06-16

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

  12. Shielding benchmark test

    International Nuclear Information System (INIS)

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  13. Combustor bulkhead heat shield assembly

    Energy Technology Data Exchange (ETDEWEB)

    Zeisser, M.H.

    1990-06-19

    This paper describes a gas turbine engine having an annular combustion chamber defined by an annular, inner liner, a concentric outer liner, and an upstream annular combustor head, wherein the head includes a radially extending bulkhead having circumferentially distributed openings for each receiving an individual fuel nozzle therethrough. It comprises: a segmented heat shield assembly, disposed between the combustion chamber interior and the bulkhead, including generally planar, sector shaped heat shields, each shield abutting circumferentially with two next adjacent shields and extending radially from proximate the inner liner to proximate the outer liner, the plurality of shields collectively defining an annular protective barrier, and wherein each sector shaped shield further includes an opening, corresponding to one of the bulkhead nozzle openings for likewise receiving the corresponding nozzle therethrough, the shield opening further including an annular lip extending toward the bulkhead and being received within the bulkhead opening, raised ridges on the shield backside, the ridges contacting the facing bulkhead surface and defining a flow path for a flow of cooling air issuing from a sized supply opening disposed in the bulkhead, the flow path running ultimately from adjacent the annular lip to the edges of each shield segment, wherein the raised edges extend fully along the lateral, circumferentially spaced edges of each shield segment and about the adjacent shield segments wherein the raised ridges further extend circumferentially between the annular lip and the abutting edge ridges.

  14. Shielding calculations for SSC

    International Nuclear Information System (INIS)

    Monte Carlo calculations of hadron and muon shielding for SSC are reviewed with emphasis on their application to radiation safety and environmental protection. Models and algorithms for simulation of hadronic and electromagnetic showers, and for production and transport of muons in the TeV regime are briefly discussed. Capabilities and limitations of these calculations are described and illustrated with a few examples. 12 refs., 3 figs

  15. Investigation of the Ge(Li) detector background in a multilayer passive shield

    International Nuclear Information System (INIS)

    The possibilities for reducing background and increasing the sensitivity of the Ge(Li) detector on the account of application of passive multilayer shield are studied. The external shield consists of concrete, polyethylene, cast iron and steel layers. The internal shield is accomplished from lead and steel-3. The results of the Ge(Li) detector background measurements in a low-background chamber by application of only one external or complete shield. The characteristics of the low-background chamber are compared with the background characteristics of other facilities. It is shown that the passive multilayer shield of the low-background chamber has advantages over the majority of facilities with single layer shield of lead

  16. 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.

  17. FIRE-RESISTANT SHIELDING COATING BASED ON SHUNGITE-CONTAINING PAINT

    OpenAIRE

    BELOUSOVA Elena Sergeevna; NASONOVA Natalia Viktorovna; LYNKOV Leonid Mihailovich; BORBOTKO Timofei Valentinovich; LISOVSKIY Dmitriy Nikolaevich

    2013-01-01

    Today when specific shielded facilities are designed the construction materials and shields should meet a range of fire safety requirements. A composite coating on the basis of a water-based fire-resistant paint filled with shungite nanopowder can be applied onto walls, floors, ceilings and other surfaces in the shielded areas to reduce electromagnetic radiation and simultaneously to ensure fire safety. Shungit is a mineral with multilayer carbon fullerene globules which diameter is 10–30 nm....

  18. The measurement technique of radiation shielding performance for hot cell walls in IMEF

    International Nuclear Information System (INIS)

    Hot cell is the facility to test irradiated materials. The capability of radiation shielding through wall should be conformed to protect the workers from expose. In this report, the measurement techniques of radiation shielding performance through hot cell walls are described. Detailed contents are as following; 1. The theory of test 2. The measuring equipment of radiations capability 3. The choice of measuring points 4. Test procedures and data analysis method 5. The reinforcement of shielding lack area. (author). 13 tabs., 19 figs

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

    CERN Document Server

    Taylor, David; Turner, Andrew; Davis, Andrew

    2014-01-01

    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").

  20. Justification for Shielded Receiver Tube Additional Lead Shielding

    International Nuclear Information System (INIS)

    In order to reduce high radiation dose rates encountered when core sampling some radioactive waste tanks the addition of 240 lbs. of lead shielding is being considered to the shielded receiver tube on core sample trucks No.1, No.3 and No.4. The lead shielding is 4 inch diameter x 1/2 inch thick half rounds that have been installed around the SR tube over its' full length. Using three unreleased but independently reviewed structural analyses HNF-6018 justifies the addition of the lead shielding

  1. Measurement of the transient shielding effectiveness of shielding cabinets

    Directory of Open Access Journals (Sweden)

    H. Herlemann

    2008-05-01

    Full Text Available Recently, new definitions of shielding effectiveness (SE for high-frequency and transient electromagnetic fields were introduced by Klinkenbusch (2005. Analytical results were shown for closed as well as for non closed cylindrical shields. In the present work, the shielding performance of different shielding cabinets is investigated by means of numerical simulations and measurements inside a fully anechoic chamber and a GTEM-cell. For the GTEM-cell-measurements, a downscaled model of the shielding cabinet is used. For the simulations, the numerical tools CONCEPT II and COMSOL MULTIPHYSICS were available. The numerical results agree well with the measurements. They can be used to interpret the behaviour of the shielding effectiveness of enclosures as function of frequency. From the measurement of the electric and magnetic fields with and without the enclosure in place, the electric and magnetic shielding effectiveness as well as the transient shielding effectiveness of the enclosure are calculated. The transient SE of four different shielding cabinets is determined and discussed.

  2. Use of MCNP in fusion blanket design ITER magnet system shielding analysis benchmark of the EFF (European Fusion File) neutron data with the FNG (Frascati Neutron Generator) 14 MeV neutron facility

    International Nuclear Information System (INIS)

    Since eight years at our laboratory, MCNP code has been used as a fundamental tool in many fusion directed activities in which we have been or we still are involved. Mainly they are: neutronics analysis of the performances of blanket components, supporting and optimizing their design; the estimation of the nuclear heat and radiation loads on the toroidal superconducting coils to assess the system shielding performances; then, a 14 MeV neutron generator is recently operating in Frascati and an experimental programme started with a benchmark neutron transport in a stainless steel block, MCNP is used to perform calculations. Present status of these experiments are reviewed. (K.A.)

  3. Passive Shielding in CUORE

    International Nuclear Information System (INIS)

    The nature of neutrino mass is one of the friontier problems of fundamental physics. Neutrinoless Double Beta Decay (0νDBD) is a powerful tool to investigate the mass hierarchy and possible extensions of the Standard Model. CUORE is a 1-Ton next generation experiment, made of 1000 Te bolometers, aiming at reaching a background of 0.01 (possibly 0.001) counts keV-1kg-1y-1 and therefore a mass sensitivity of few tens of meV The background contribution due to environmental neutrons, muon-induced neutrons in the shieldings and external gamma is discussed

  4. Planar Shielded-Loop Resonators

    OpenAIRE

    Tierney, Brian B.; Grbic, Anthony

    2014-01-01

    The design and analysis of planar shielded-loop resonators for use in wireless non-radiative power transfer systems is presented. The difficulties associated with coaxial shielded-loop resonators for wireless power transfer are discussed and planar alternatives are proposed. The currents along these planar structures are analyzed and first-order design equations are presented in the form of a circuit model. In addition, the planar structures are simulated and fabricated. Planar shielded-loop ...

  5. Walls shielding against ionizing radiation

    International Nuclear Information System (INIS)

    These specifications are to help the users of lead bricks as under DIN 25407, leaf 1, with the construction of walls shielding against ionizing radiation by examples for the uses of the different types of lead bricks and by recommendations for the construction of shielding walls and for the determination of the wall thickness necessary for shielding against γ-radiation as a function of energy. (orig./AK)

  6. SHIELD verification and validation report

    Energy Technology Data Exchange (ETDEWEB)

    Boman, C.

    1992-02-01

    This document outlines the verification and validation effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system code. Along with its predecessors, SHIELD has been in use at the Savannah River Site (SRS) for more than ten years. During this time the code has been extensively tested and a variety of validation documents have been issued. The primary function of this report is to specify the features and capabilities for which SHIELD is to be considered validated, and to reference the documents that establish the validation.

  7. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4He 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

  8. New Toroid shielding design

    CERN Multimedia

    Hedberg V

    On the 15th of June 2001 the EB approved a new conceptual design for the toroid shield. In the old design, shown in the left part of the figure above, the moderator part of the shielding (JTV) was situated both in the warm and cold areas of the forward toroid. It consisted both of rings of polyethylene and hundreds of blocks of polyethylene (or an epoxy resin) inside the toroid vacuum vessel. In the new design, shown to the right in the figure above, only the rings remain inside the toroid. To compensate for the loss of moderator in the toroid, the copper plug (JTT) has been reduced in radius so that a layer of borated polyethylene can be placed around it (see figure below). The new design gives significant cost-savings and is easier to produce in the tight time schedule of the forward toroid. Since the amount of copper is reduced the weight that has to be carried by the toroid is also reduced. Outgassing into the toroid vacuum was a potential problem in the old design and this is now avoided. The main ...

  9. Iron shielded MRI optimization

    Science.gov (United States)

    Borghi, C. A.; Fabbri, M.

    1998-09-01

    The design of the main current systems of an actively shielded and of an iron shielded MRI device for nuclear resonance imaging, is considered. The model for the analysis of the magnetic induction produced by the current system, is based on the combination of a Boundary Element technique and of the integration of two Fredholm integral equations of the first and the second kind. The equivalent current magnetization model is used for the calculation of the magnetization produced by the iron shield. High field uniformity in a spherical region inside the device, and a low stray field in the neighborhood of the device are required. In order to meet the design requirements a multi-objective global minimization problem is solved. The minimization method is based on the combination of the filled function technique and the (1+1) evolution strategy algorithm. The multi-objective problem is treated by means of a penalty method. The actively shielded MRI system results to utilize larger amount of conductor and produce higher magnetic energy than the iron shield device. On veut étudier le projet du système des courants principaux d'un MRI à écran en fer et d'un MRI à écran actif. Le modèle d'analyse du champ magnétique produit par le système de courants est basé sur la combinaison d'une technique Boundary Element et de l'intégration de deux équations intégrales de Fredholm de première et de seconde sorte. On utilise pour calculer la magnétisation produite par l'écran en fer le modèle à cou rants de magné ti sa tion équivalents. On exige une élévation uniforme du champ dans une région sphérique au cœur de l'appareil et un bas champ magnétique dispersé à proximité de l'appareil. Dans le but de répondre aux impératifs du projet, on va résoudre un problème multiobjectif de minimisation globale. On utilise une technique de minimisation obtenue par la combinaison des méthodes “Filled Function” et “(1+1) Evolution Strategy”. Le probl

  10. 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. PMID:21366239

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

    International Nuclear Information System (INIS)

    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)

  12. Drip Shield Emplacement Gantry Concept

    Energy Technology Data Exchange (ETDEWEB)

    Silva, R.A.; Cron, J.

    2000-03-29

    This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existing equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made

  13. Analysis of some splitting and roulette algorithms in shield calculations by the Monte Carlo method

    International Nuclear Information System (INIS)

    Different schemes of using the splitting and roulette methods in calculation of radiation transport in nuclear facility shields by the Monte Carlo method are considered. Efficiency of the considered schemes is estimated on the example of test calculations

  14. Development of superconductor bulk for superconductor bearing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chan Joong; Jun, Byung Hyuk; Park, Soon Dong (and others)

    2008-08-15

    Current carrying capacity is one of the most important issues in the consideration of superconductor bulk materials for engineering applications. There are numerous applications of Y-Ba-Cu-O (YBCO) bulk superconductors e.g. magnetic levitation train, flywheel energy storage system, levitation transportation, lunar telescope, centrifugal device, magnetic shielding materials, bulk magnets etc. Accordingly, to obtain YBCO materials in the form of large, single crystals without weak-link problem is necessary. A top seeded melt growth (TSMG) process was used to fabricate single crystal YBCO bulk superconductors. The seeded and infiltration growth (IG) technique was also very promising method for the synthesis of large, single-grain YBCO bulk superconductors with good superconducting properties. 5 wt.% Ag doped Y211 green compacts were sintered at 900 .deg. C {approx} 1200 .deg.C and then a single crystal YBCO was fabricated by an infiltration method. A refinement and uniform distribution of the Y211 particles in the Y123 matrix were achieved by sintering the Ag-doped samples. This enhancement of the critical current density was ascribable to a fine dispersion of the Y211 particles, a low porosity and the presence of Ag particles. In addition, we have designed and manufactured large YBCO single domain with levitation force of 10-13 kg/cm{sup 2} using TSMG processing technique.

  15. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    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 PuO2 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

  16. Shielding measurements for a 230 MeV proton beam

    International Nuclear Information System (INIS)

    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 cm3 to 1000 cm3 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

  17. The shield effect

    DEFF Research Database (Denmark)

    Toft, Søren; Albo, Maria J

    2016-01-01

    Several not mutually exclusive functions have been ascribed to nuptial gifts across different taxa. Although the idea that a nuptial prey gift may protect the male from pre-copulatory sexual cannibalism is attractive, it has previously been considered of no importance based on indirect evidence and...... rejected by experimental tests. We reinvestigated whether nuptial gifts may function as a shield against female attacks during mating encounters in the spider Pisaura mirabilis and whether female hunger influences the likelihood of cannibalistic attacks. The results showed that pre-copulatory sexual...... cannibalism was enhanced when males courted without a gift and this was independent of female hunger. We propose that the nuptial gift trait has evolved partly as a counteradaptation to female aggression in this spider species....

  18. Aladdin upgrade design study: shielding

    International Nuclear Information System (INIS)

    The object of this shielding is to examine all aspects of Aladdin operation to ensure that adequate shielding is provided to meet the design objectives. To do this, we will look at shielding necessary for radiation produced during the injection process, during normal loss of the stored beam and during accidental loss of the stored beam. It will therefore be necessary to specify shielding not only at the ring, but also along the injection line and the optical beam lines. We will also give special attention to the occupation of the accelerator Vault during injection as this may be a desirable design option. In effect, two shielding plans will be presented, permitting estimates of cost and space requirements for both

  19. Welding shield for coupling heaters

    Science.gov (United States)

    Menotti, James Louis

    2010-03-09

    Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

  20. Parameters calculation of shielding experiment

    International Nuclear Information System (INIS)

    The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author)

  1. Effectiveness of custom neutron shielding in the maze of radiotherapy accelerators

    International Nuclear Information System (INIS)

    An investigation was performed to examine the neutron dose equivalent in a radiotherapy maze lined with a customised neutron shielding material. The accelerator investigated was a Varian Clinac 2100C/D using 18 MV photons, and the neutron shielding utilised at this centre was PremadexTM commercially available neutron shielding. Based on Monte Carlo simulations, properly installed customised neutron shielding may reduce the neutron dose equivalent by up to a factor of 8 outside the maze, depending upon the installation. In addition, it was determined that the neutron dose near the entrance to the maze may be reduced by approximately 40% by using customised neutron shielding in the maze, as compared with a facility not using this shielding. This would have a positive dose-saving effect in doorless maze designs. (author)

  2. Shielding benchmark experiments through concrete and iron with high-energy proton and heavy ion accelerators

    International Nuclear Information System (INIS)

    The deep penetration of neutrons through thick shield has become a very serious problem in the shielding design of high-energy, high-intensity accelerator facility. In the design calculation, the Monte Carlo transport calculation through thick shields has large statistical errors and the basic nuclear data and model used in the existing Monte Carlo codes are not well evaluated because of very few experimental data. It is therefore strongly needed to do the deep penetration experiment as shielding benchmark for investigating the calculation accuracy. Under this circumference, we performed the following two shielding experiments through concrete and iron, one with a 800 MeV proton accelerator of the rutherford appleton laboratory (RAL), England and the other with a high energy heavy iron accelerator of the national institute of radiological sciences (NIRS), Japan. Here these two shielding benchmark experiments are outlined. (orig.)

  3. Shielding benchmark experiments through concrete and iron with high-energy proton and heavy ion accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, T.; Sasaki, M.; Nunomiya, T.; Iwase, H. [Tohoku Univ., Sendai (Japan). Dept. of Quantum Science and Energy Engineering; Nakao, N.; Shibata, T. [High Energy Accelerator Research Organization (KEK), Ibaraki (Japan); Kim, E. [Japan Atomic Energy Research Inst. (JAERI), Ibaraki (Japan). Tokai Establishment; Kurosawa, T. [Japan Synchrotron Radiation Research Inst. (JASRI), Hyogo (Japan); Taniguchi, S. [Electrotechnical Lab. (ETL), Tsukuba, Ibaraki (Japan); Uwamino, Y.; Ito, S. [The Inst. of Physical and Chemical Research (RIKEN), Saitama (Japan); Fukumura, A. [National Inst. of Radiological Sciences (NIRS), Chiba (Japan); Perry, D.R.; Wright, P. [Rutherford Appleton Lab. (RAL), Didcot, Oxfordshire (United Kingdom). Health and Safety Group

    2001-07-01

    The deep penetration of neutrons through thick shield has become a very serious problem in the shielding design of high-energy, high-intensity accelerator facility. In the design calculation, the Monte Carlo transport calculation through thick shields has large statistical errors and the basic nuclear data and model used in the existing Monte Carlo codes are not well evaluated because of very few experimental data. It is therefore strongly needed to do the deep penetration experiment as shielding benchmark for investigating the calculation accuracy. Under this circumference, we performed the following two shielding experiments through concrete and iron, one with a 800 MeV proton accelerator of the rutherford appleton laboratory (RAL), England and the other with a high energy heavy iron accelerator of the national institute of radiological sciences (NIRS), Japan. Here these two shielding benchmark experiments are outlined. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Magnetic shielding properties of a superconducting hollow cylinder containing slits: Modelling and experiment

    OpenAIRE

    Fagnard, Jean-François; Elschner, S.; Hobl, A.; Bock, J.; Vanderheyden, Benoît; Vanderbemden, Philippe

    2012-01-01

    This paper deals with the magnetic properties of bulk high temperature superconducting cylinders used as magnetic shields. We investigate, both numerically and experimentally, the magnetic properties of a hollow cylinder with two axial slits which cut the cylinder in equal halves. Finite element method modelling has been used with a three-dimensional geometry to help us in understanding how the superconducting currents flow in such a cut cylinder and therefore how the magnetic shielding prope...

  6. Soil density and mass attenuation coefficients for use in shielding calculations at the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Compacted, backfilled soil excavated during construction may be used to provide shielding from gamma radiation at the Hanford Waste Vitrification Plant (HWVP). To provide a reasonable estimate of the shielding offered by this backfilled soil, the bulk density and the composition of the emplaced soil must be specified. This study provides an estimate of the bulk density and the mass attenuation coefficients of soil used for calculating gamma-ray shielding attenuation at the HWVP. These estimates are based on measurements taken from soil samples and underlying rock samples at the Hanford Site

  7. Neutron shielding evaluation for a small fuel transport case

    CERN Document Server

    Coeck, M; Vanhavere, F

    2002-01-01

    We investigated the effectiveness of a small neutron shield configuration for the transportation of fresh MOX fuel rods in an experimental facility, this in order to reduce the dose received by the personnel. Monte Carlo simulations using the Tripoli and MCNP4B code were applied. Different configurations were studied, starting from the bare fuel rod positioned on an iron plate up to a fuel rod covered by a box-shaped shield made of different materials such as polyethylene, polyethylene with boron and polyethylene with a cadmium layer. We compared the neutron spectra for the different cases and calculated the corresponding ambient equivalent dose rate H*(10).

  8. System for imaging plutonium through heavy shielding

    International Nuclear Information System (INIS)

    A single pinhole can be used to image strong self-luminescent gamma-ray sources such as plutonium on gamma scintillation (Anger) cameras. However, if the source is weak or heavily shielded, a poor signal to noise ratio can prevent acquisition of the image. An imaging system designed and built at Los Alamos National Laboratory uses a coded aperture to image heavily shielded sources. The paper summarizes the mathematical techniques, based on the Fast Delta Hadamard transform, used to decode raw images. Practical design considerations such as the phase of the uniformly redundant aperture and the encoded image sampling are discussed. The imaging system consists of a custom designed m-sequence coded aperture, a Picker International Corporation gamma scintillation camera, a LeCroy 3500 data acquisition system, and custom imaging software. The paper considers two sources - 1.5 mCi 57Co unshielded at a distance of 27 m and 220 g of bulk plutonium (11.8% 240Pu) with 0.3 cm lead, 2.5 cm steel, and 10 cm of dense plastic material at a distance of 77.5 cm. Results show that the location and geometry of a source hidden in a large sealed package can be determined without having to open the package. 6 references, 4 figures

  9. Radiation protection effectiveness of a proposed magnetic shielding concept for manned Mars missions

    Science.gov (United States)

    Townsend, Lawrence W.; Wilson, John W.; Shinn, J. L.; Nealy, John E.; Simonsen, Lisa C.

    1990-01-01

    The effectiveness of a proposed concept for shielding a manned Mars vehicle using a confined magnetic field configuration is evaluated by computing estimated crew radiation exposures resulting from galactic cosmic rays and a large solar flare event. In the study the incident radiation spectra are transported through the spacecraft structure/magnetic shield using the deterministic space radiation transport computer codes developed at Langley Research Center. The calculated exposures unequivocally demonstrate that magnetic shielding could provide an effective barrier against solar flare protons but is virtually transparent to the more energetic galactic cosmic rays. It is then demonstrated that through proper selection of materials and shield configuration, adequate and reliable bulk material shielding can be provided for the same total mass as needed to generate and support the more risky magnetic field configuration.

  10. Radiation shields for a shelter

    International Nuclear Information System (INIS)

    A simple and cheap closure and radiation shield arrangement is described for the entrance of an underground shelter. The shelter can serve as a blast-proof, biological or nuclear shelter. The radiation shield is positioned above the habitable space of the shelter and below a blast-proof, dust-proof outer cover. The shield consists of a box containing a filling, e.g. coke with a concrete screed, is closed by bolted panels and is horizontally moveable by sliding on castors. (author)

  11. New Materials for EMI Shielding

    Science.gov (United States)

    Gaier, James R.

    1999-01-01

    Graphite fibers intercalated with bromine or similar mixed halogen compounds have substantially lower resistivity than their pristine counterparts, and thus should exhibit higher shielding effectiveness against electromagnetic interference. The mechanical and thermal properties are nearly unaffected, and the shielding of high energy x-rays and gamma rays is substantially increased. Characterization of the resistivity of the composite materials is subtle, but it is clear that the composite resistivity is substantially lowered. Shielding effectiveness calculations utilizing a simple rule of mixtures model yields results that are consistent with available data on these materials.

  12. Noise Shielding Using Acoustic Metamaterials

    International Nuclear Information System (INIS)

    We exploit theoretically a class of rectangular cylindrical devices for noise shielding by using acoustic metamaterials. The function of noise shielding is justified by both the far-field and near-field full-wave simulations based on the finite element method. The enlargement of equivalent acoustic scattering cross sections is revealed to be the physical mechanism for this function. This work makes it possible to design a window with both noise shielding and air flow. (electromagnetism, optics, acoustics, heat transfer, classical mechanics, and fluid dynamics)

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

    Science.gov (United States)

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

    2016-08-01

    Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam 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.

  14. Photon attenuation characteristics of radiation shielding materials

    International Nuclear Information System (INIS)

    surface. The thickness of the absorber was increased in steps by adding absorbers as indicated earlier. From the transmitted (I) and the incident photon intensity (I0), for a thickness 'x' of the absorber, the photon attenuation coefficient μm is calculated by the following expression: μ=ln(l0/l)X. The values of photon attenuation coefficients thus are obtained. The overall uncertainty of the measured values was estimated to be around 5% and had the following components: the counting statistics for I and I0 measurements and thickness uniformity of the absorbers. However, as the data for commercial cement and barite are not available, no comparison have been made with the theoretical values. Thus the derived values of attenuation coefficients of the tested materials can be utilised in compilation of shielding thicknesses for any radiation facilities where these materials are to be used. (author)

  15. AP600 Shield building

    International Nuclear Information System (INIS)

    In order to minimize capital costs and save time in the global construction time schedule for the AP600 Nuclear Power Plant, planned in 36 months from excavation up to the fuel charging, ANSALDO has developed an innovative Shield Building Conical Roof design having the following basic characteristics: i) can be erected approximately in less than two months; ii) allows the functionality of the Passive Containment Cooling System (PCSS) located in the PCCS tank and in the Valve Room anchored directly to the conical roof itself; iii) satisfies the structural loads design as Safe Shutdown Earthquake, or the Aircraft Crash and both integrated with the sloshing analysis for the tank located at the top of the conical roof. The most important aspects of this new roof are: a) use of prefabricated precast panels; b) address the erection of the formworks using temporary structures having the capability of becoming final elements; c) develop a modular rebars sizing and design in order to perform the most important portion of the job in the workshop; d) second pouring construction sequence assuring full integration with the formwork function; e) modular construction of the PCSS tank at the top of the conical roof. An interesting evaluation has been also performed in calculating sloshing phenomenon in the PCSS tank by comparing detailed 3D Finite Element Model approach and simplified qualified formulas dedicated to this phenomenon. (author). 2 figs

  16. Shielding of moving line charges

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Youmei; He, Bingyu [Department of Physics, School of Science, Hangzhou Dianzi University, Hangzhou 310018 (China); Yu, Wei [Shanghai Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Shanghai 201800 (China); Yu, M.Y., E-mail: myyu@zju.edu.cn [Institute for Fusion Theory and Simulation and Department of Physics, Zhejiang University, Hangzhou 310027 (China); Institute for Theoretical Physics I, Ruhr University, D-44780 Bochum (Germany)

    2015-07-03

    A charged object moving in plasma can excite plasma waves that inevitably modify its Debye shielding characteristics. When the excited waves propagate sufficiently fast, the shielding can even break down. Here the properties of finite amplitude plasma waves excited by a moving line charge are investigated. It is found that when the speed of the latter is close to but less than the thermal speed of the background plasma electrons, only a localized disturbance in the form of a soliton that moves together with the line charge is excited. That is, the line charge is well shielded even though it is moving at a high speed and has generated a large local electrostatic field. However, for a pair of line charges moving together, such complete shielding behavior could not be found.

  17. Performance test on shielding concrete

    International Nuclear Information System (INIS)

    The cylinder of the shielding concrete is made from common Portland cement and home-made coarse or fine aggregates. Orthogonal design experiment and regression analysis are adopted to study the effects of the water content, sand percentage and water-cement ratio on the property of shielding concrete and the difference between them. The test shows that the tensile strength is in inverse proportion with water-cement ratio, and the influence is quite significant. Another factor is the type of aggregates. The effect of the age on its density is not obvious. Similarly, the concrete shielding γ rays shares the same influencing factors with that shielding neutron rays on density, slump and tensile strength. And both have the same change rules regarding to mechanical property. (authors)

  18. Concrete mix design for X-and gamma shielding

    International Nuclear Information System (INIS)

    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)

  19. International Space Station (ISS) Meteoroid/Orbital Debris Shielding

    Science.gov (United States)

    Christiansen, Eric L.

    1999-01-01

    Design practices to provide protection for International Space Station (ISS) crew and critical equipment from meteoroid and orbital debris (M/OD) Impacts have been developed. Damage modes and failure criteria are defined for each spacecraft system. Hypervolocity Impact -1 - and analyses are used to develop ballistic limit equations (BLEs) for each exposed spacecraft system. BLEs define Impact particle sizes that result in threshold failure of a particular spacecraft system as a function of Impact velocity, angles and particle density. The BUMPER computer code Is used to determine the probability of no penetration (PNP) that falls the spacecraft shielding based on NASA standard meteoroid/debris models, a spacecraft geometry model, and the BLEs. BUMPER results are used to verify spacecraft shielding requirements Low-weight, high-performance shielding alternatives have been developed at the NASA Johnson Space Center (JSC) Hypervelocity Impact Technology Facility (HITF) to meet spacecraft protection requirements.

  20. Shielding evaluation by laser compton scattering gamma-ray

    International Nuclear Information System (INIS)

    Laser Compton scattering gamma-ray beam was used for evaluation of gamma ray shield. The gamma source of a NewSUBARU Synchrotron Radiation Facility can generate the quasi-monochromatic gamma ray beam of 0.5-1.7 MeV by combining a carbon dioxide laser and a 0.5-1.0 GeV electron beam. This gamma-ray source has small divergence of 1/γ radian due to the relativistic effect, where γ is relativistic factor of electron. Small diameter test beam of gamma-ray of about 1 mm in diameter is possible to use at the 10 m from the gamma-ray source by combining the small divergence gamma-ray beam with small hole lead collimator. Test sample size used was 2 cm in diameter. Measured shield factor was compared with calculated value using known shield materials such as lead. (author)

  1. Radiation shield for LHCb experiment at Point 8

    CERN Document Server

    Lacarrère, D; Lindner, Rolf

    1998-01-01

    This report describes the conceptual design of the main shielding wall which is required for protecting against radiation the counting rooms of LHCb in the underground experimental area (UX85) at Point 8. The radiation simulations have been performed using the Monte Carlo code FLUKA97. The results, defining the main characteristics of the shield are presented. The major implications to the existing facilities and services, generated from the erection of the wall, dividing the existing cavern in the two areas, are investigated and the engineering solutions for the modifications are explored. The most appropriate methods of construction for erecting the shielding wall are detailed and compared. Preliminary cost estimates and planning are also mentioned.1

  2. Transparent nanostructured coatings with UV-shielding and superhydrophobicity properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang Taoye; Chen Jianfeng [Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Isimjan, Tayirjan T; Rohani, Sohrab, E-mail: chenjf@mail.buct.edu.cn, E-mail: srohani@uwo.ca [Department of Chemical and Biochemical Engineering, University of Western Ontario, 1151 Richmond Street, London, ON, N6A 3K7 (Canada)

    2011-07-01

    Visible light transparent, UV-shielding and superhydrophobic nanostructured coatings have been successfully fabricated through a facile layer-by-layer deposition of TiO{sub 2} and SiO{sub 2} nanoparticles. The coatings are composed of an underlying UV-shielding TiO{sub 2} layer and a top fully covered protective SiO{sub 2} layer. The resulting coatings can block 100% of UVB and UVC and almost 85% of UVA. The fabricated surfaces have contact angles exceeding 165 deg. after coating with organic PTES (1H, 1H, 2H, 2H-perfluorooctyltriethoxysilane) molecules. The transparent superhydrophobic surfaces exhibit extremely strong UV stability. All coatings retain the initial UV-shielding and superhydrophobic properties even after exposure to 275 nm UV light with a light intensity of 75 mW cm{sup -2} for 12 h.

  3. Shielding vacuum fluctuations with graphene

    OpenAIRE

    Ribeiro, Sofia; Scheel, Stefan

    2013-01-01

    The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are...

  4. Composite Aerogel Multifoil Protective Shielding

    Science.gov (United States)

    Jones, Steven M.

    2013-01-01

    New technologies are needed to survive the temperatures, radiation, and hypervelocity particles that exploration spacecraft encounter. Multilayer insulations (MLIs) have been used on many spacecraft as thermal insulation. Other materials and composites have been used as micrometeorite shielding or radiation shielding. However, no material composite has been developed and employed as a combined thermal insulation, micrometeorite, and radiation shielding. By replacing the scrims that have been used to separate the foil layers in MLIs with various aerogels, and by using a variety of different metal foils, the overall protective performance of MLIs can be greatly expanded to act as thermal insulation, radiation shielding, and hypervelocity particle shielding. Aerogels are highly porous, low-density solids that are produced by the gelation of metal alkoxides and supercritical drying. Aerogels have been flown in NASA missions as a hypervelocity particle capture medium (Stardust) and as thermal insulation (2003 MER). Composite aerogel multifoil protective shielding would be used to provide thermal insulation, while also shielding spacecraft or components from radiation and hypervelocity particle impacts. Multiple layers of foil separated by aerogel would act as a thermal barrier by preventing the transport of heat energy through the composite. The silica aerogel would act as a convective and conductive thermal barrier, while the titania powder and metal foils would absorb and reflect the radiative heat. It would also capture small hypervelocity particles, such as micrometeorites, since it would be a stuffed, multi-shock Whipple shield. The metal foil layers would slow and break up the impacting particles, while the aerogel layers would convert the kinetic energy of the particles to thermal and mechanical energy and stop the particles.

  5. Shielding requirements in helical tomotherapy

    International Nuclear Information System (INIS)

    Helical tomotherapy is a relatively new intensity-modulated radiation therapy (IMRT) treatment for which room shielding has to be reassessed for the following reasons. The beam-on-time needed to deliver a given target dose is increased and leads to a weekly workload of typically one order of magnitude higher than that for conventional radiation therapy. The special configuration of tomotherapy units does not allow the use of standard shielding calculation methods. A conventional linear accelerator must be shielded for primary, leakage and scatter photon radiations. For tomotherapy, primary radiation is no longer the main shielding issue since a beam stop is mounted on the gantry directly opposite the source. On the other hand, due to the longer irradiation time, the accelerator head leakage becomes a major concern. An analytical model based on geometric considerations has been developed to determine leakage radiation levels throughout the room for continuous gantry rotation. Compared to leakage radiation, scatter radiation is a minor contribution. Since tomotherapy units operate at a nominal energy of 6 MV, neutron production is negligible. This work proposes a synthetic and conservative model for calculating shielding requirements for the Hi-Art II TomoTherapy unit. Finally, the required concrete shielding thickness is given for different positions of interest

  6. Integral Face Shield Concept for Firefighter's Helmet

    Science.gov (United States)

    Abeles, F.; Hansberry, E.; Himel, V.

    1982-01-01

    Stowable face shield could be made integral part of helmet worn by firefighters. Shield, made from same tough clear plastic as removable face shields presently used, would be pivoted at temples to slide up inside helmet when not needed. Stowable face shield, being stored in helmet, is always available, ready for use, and is protected when not being used.

  7. Transient heat flux shielding using thermal metamaterials

    Science.gov (United States)

    Narayana, Supradeep; Savo, Salvatore; Sato, Yuki

    2013-05-01

    We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

  8. Transient heat flux shielding using thermal metamaterials

    CERN Document Server

    Narayana, Supradeep; Sato, Yuki

    2013-01-01

    We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

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

    International Nuclear Information System (INIS)

    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

  10. PRESTO, Slab Shields for Time-Dependent Gamma Spectra

    International Nuclear Information System (INIS)

    A - Description of program or function: PRESTO is designed to calculate slab shields for gamma ray sources of complex and time dependent energy spectra. PRESTO I treats cylinder sources with shields at side, such as pipelines or containers in radioactive facilities. PRESTO II is the analogous code for spherical sources. The programs permit to consider volume sources or a combination of volume and surface sources. To describe the source spectrum, one can start from the nuclides contained in the source mixture or (with the aid of PRESTO IA) from energy group sets. The internal data set contains 5 common shield construction materials. B - Method of solution: The solution method is based on the point kernel integration, extended by the 'self absorption distance' concept. The approximation mentioned before reduces the spatial flux calculation to a plane problem. The dose build-up factor is taken into account by Taylor's equation. Some functions necessary for the integration will be calculated inside the program, by means of extrapolation based on the internal data sets. PRESTO permits to calculate: 1). required shield thicknesses for a given dose rate level or the allowed activity concentration of the source for a given shield thickness, both time dependent. 2). the contribution to the dose rate by single nuclides. C - Restrictions on the complexity of the problem: - Energy range from 0.1 to 10 MeV. - Contents of internal nuclide library: 100 nuclides with 428 energy lines. One job run can use 30 of them (up to 280 energy lines) - PRESTO I : 1 shielding material per job - PRESTO II: up to 5 shielding materials per job - PRESTO IA: 40 energy groups, taken from the nuclides contained in the data set, or up to 250 free energy groups

  11. Study on the development and application of high-performance shielding materials of polymer-type

    International Nuclear Information System (INIS)

    It is necessary to make the shield structure of marine reactor lighter and smaller until nuclear ships are put into practical use. Meanwhile, efforts have been made to reduce the exposure of employees in nuclear facilities following instructions based on the ALARA spirit, which requires more shields. Development of high-performance shielding materials is one of the ways to meet these requirements which oppose each other. Such materials are also very useful for the transport or the storage technology of radioactive materials. Studies are carried out to develop high-performance shielding materials that have good shielding performance for both neutrons and gamma rays and have low activation property. As raw materials, lead and boron or their chemical compounds and high hydrogen content monomers are selected. They are mixed and made into a solid state by using an ultraviolet curing system. Computational analysis of the materials proves their shielding and activation properties sufficiently well as expected. That is, an ordinary concrete shield can be replaced with the material of about half thickness. Furthermore, almost no changes observed in their weight when they are heated up to 100degC degree from room temperature. Tentatively, one of the materials is applied as a compensational shield in irregularities, that is, a bent duct and an annular offset gap. Calculations prove that the application of the material to these shield defects is made successfully. (author)

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

    International Nuclear Information System (INIS)

    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.

  13. Novel light-weight materials for shielding gamma ray

    Science.gov (United States)

    Chen, Shuo; Bourham, Mohamed; Rabiei, Afsaneh

    2014-03-01

    A comparison of gamma ray attenuation effectiveness of bulk aluminum, close-cell composite metal foams and open-cell aluminum foam infiltrated with variety of second phase materials were investigated and reported in this study. Mass attenuation coefficients for six sets of samples with three different areal densities of 2, 5 and 10 g/cm2 were determined at photon energies of 0.060, 0.662, 1.173, and 1.332 MeV. Theoretical values were calculated using XCOM software package. A complete agreement was observed between experimental and theoretical results. It is observed that close-cell composite metal foams exhibit a better shielding capability compared to open-cell Al foam with fillers. It is also observed that close-cell composite metal foams offer superior shielding effectiveness compared to bulk aluminum for energies below 0.662 MeV, the mass attenuation coefficients of steel-steel composite metal foam and Al-steel composite metal foam were measured 400 and 300% higher than that of aluminum A356. This study indicates the potential of utilizing the light-weight composite metal foams as shielding material replacing current heavy materials used for attenuation of low energy gamma ray with additional advantages such as high energy absorption and excellent heat rejection capabilities.

  14. Neutron shielding calculation for VVER NPP

    International Nuclear Information System (INIS)

    There are two methods for neutron transport (shielding) calculation used in Energoproject, Prague, the method of discrete ordinates (code TORT-DORT) and the Monte Carlo method (codes MCNP and module within the code SCALE). The task concerning neutron dose rates calculation near casks with VVER spent fuel are presented as an example. Measured neutron dose rates of real loaded C-30 casks for VVER spent fuel assemblies are compared with calculated values in the frame of the international benchmark calculation task. A part of the task realized by the Atomic Energy Research (AER) organization concerning neutron shielding is calculated. The cask C-30 is used in Slovak Jaslovske Bohunice NPP for transport of spent fuel assemblies to the storage facility. The benchmark task has been calculated by the two-dimensional code DORT originated from Oak Ridge National Laboratory. The code solves transport problems using the method of discrete ordinates (SN - method). Calculated neutron dose rates in azimuth and vertical directions show good agreement with the experiment within the range of the measurement errors. In comparison with the other codes the results of DORT are approximately 20% lower. There have been analysed differences between one- and two- dimensional approach and influence of the flux-to-dose rate conversion factors set

  15. Neutron shielding of the GDT (Novosibirsk) neutron source project: A feasibility study

    International Nuclear Information System (INIS)

    The paper presents results of extensive neutronic studies of the neutron source test facility based on the Novosibirsk gas dynamic trap. The facility is to provide 1018 DT-neutrons/s for material-test studies. The paper examines the protective-shields capacity to ensure survival of GDT vital parts and suggests design modifications when survival is in jeopardy. The numerical studies used the 3D-AMC-VINIA Monte Carlo code with a precise computer representation of the sensitive parts of the facility. Shielding feasibility has been ascertained, and the lifetime of consumable components ensured beyond the recommended values

  16. Shielding analysis for ITER equatorial bio-shield plug

    International Nuclear Information System (INIS)

    ITER equatorial port cell outside bio-shield plug is a place for allowing free personnel access after shutdown which accommodates various sensitive equipment and pipes. To ensure the personnel safety in port cell after shutdown, the distribution of dose rate in port cell was studied. Based on VisualBUS (CAD-Based Multi-Functional 4D Neutronics Simulation System), dose rate calculations were completed in port cell after shutdown. The result showed that dose rates in port cell are still 2 orders of magnitude more than desired limit (10 μSv/h) after one day shutdown. The optimization of bio-shield was needed. (authors)

  17. Shielding and Activation Analyses in Support of the Spallation Neutron Source (SNS) ES and H Requirements

    International Nuclear Information System (INIS)

    Shielding and activation analyses play an important part in determining how to meet the Environmental, Safety and Health (ES and H) requirements of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS). The shielding and activation analyses described in this paper were performed primarily using the CALOR code system coupled with MCNP for radiation transport, the ORIHET95 isotope generation and depletion code for activation analysis, and the DOORS multi-dimensional discrete ordinates transport code system for shielding analyses. Additionally, a portion of the shielding calculations were performed with the semi-empirical code - CASL. This paper gives an overview of relevant ES and H policies and requirements, and provides detailed discussions of the shielding and activation analyses completed in support of those policies and requirements

  18. DUPIC facility engineering

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. MMW [multimegawatt] shielding design and analysis

    International Nuclear Information System (INIS)

    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

  1. Radiation Shielding Materials and Containers Incorporating Same

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-01

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

  2. 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.

  3. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

  4. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

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

  5. Some benchmark shielding problems solved by the finite element method

    International Nuclear Information System (INIS)

    Some of the test cases on bulk shields for the two-dimensional codes MARC, TRIMOM and FELICIT are described. These codes use spherical harmonic expansions for neutron directions and a finite element grid over space. MARC was developed primarily as a reactor physics code with a finite element option and it assumes isotropic scattering. TRIMOM is being developed as a general purpose shielding code for anisotropic scatterers. FELICIT is being developed as a module of TRIMOM for cylindrical systems. All three codes employ continuous trial functions at present. Exploratory work on the use of discontinuous trial functions is described. Discontinuous trial functions permit the splicing of methods which use different angular expansions, so that, for example, transport theory can be used where it is necessary and diffusion theory can be used elsewhere. (author)

  6. Cask storing facility

    International Nuclear Information System (INIS)

    The present invention provides a facility suitable to keeping and storing of casks for transporting and storing spent fuels generated from power plants and radioactive wastes generated from spent fuel reprocessing plants. Namely, the casks are transported in and out by a portal crane when they are stored. The cask storage space is disposed underground and soils are used as a portion of shielding materials. Then, a portal crane gives less load on the storage building when it is used compared with a case of using an overhead traveling crane. Since the storage pits are disposed underground, the radiation released from the casks in lateral and downward directions can be shielded by the soils. If shielding lids are disposed on the upper portion of the cask storage pits, upward radiation released from the casks can be shielded. Accordingly, there is no need to ensure thickness of walls of the building and ceilings for shielding. As a result, construction cost for the building can be reduced. (I.S.)

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

    International Nuclear Information System (INIS)

    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

  8. 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

  9. Survivor shielding. Part B. Improvements in building shielding

    International Nuclear Information System (INIS)

    Most atomic-bomb survivor doses are affected by the shielding provided by wooden structures, either in which the survivor resides or which lie between him or her and the epicenter. In the dosimetry system, this shielding of survivors can be described by a transmission factor (TF), which is the ratio of the dose with and without the structures being present. The TF typically ranges between 0.3 and 1.0. After DS86 was implemented at RERF, several of the shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, a large bimodal uncertainty in the 9-parameter category 'FS=0' was identified. Corrective action was proposed and is now implemented in DS02. In 2002, a dose bias in large wooden buildings, such as schools, was identified and a correction is implemented in DS02. A correction is also implemented in DS02 to take care of a large uncertainty in the globe-house shielding. (J.P.N.)

  10. Manufacture of a shield prototype for primary wall modules

    International Nuclear Information System (INIS)

    In the frame of the BLANKET MODULE (BM) development for ITER, an R and D programme was implemented for the manufacture of a shield prototype by powder Hot Isostatic Pressing (HIPping). The manufactured shield is a full scale module No. 11a. Starting from a forged block of 1200 x 1200 x 500 mm, the main machining steps as deep drilling (1200 mm), 3D machining and sawing were performed. Tubes were 3D bent and large number of small parts were designed and machined. By welding together all the sub-parts we erected the main part of the water coolant circuit. Once the water circuit was built; the shield was completed using powder HIPping together with forged block embedding the tubes and their in a final solid part. The powder/solid HIP is used to minimize the number of BM seal welds in front of plasma. It increases the reliability of the components during operation. About 300 kg of stainless steel powder was densified together with the forged block. 3D measurement was done before and after the HIP cycle to collect the data to be compared with theoretical model. It allows to predict the main distortions of the solid bulk. Ultrasonic examination of the densified powder on the Stainless steel bulk and around the bended tubes was performed as well as mechanical characterization of the samples. The recess for stub key attachment on the vacuum vessel side, the hydraulic connector, the key for the primary wall panel attachment on the front side and the link between the four parallel water coolant circuits were then machined to achieve the shield prototype. (orig.)

  11. Manufacture of a shield prototype for primary wall modules

    International Nuclear Information System (INIS)

    In the frame of the blanket module (BM) development for ITER, an R and D programme was implemented for the manufacture of a shield prototype by powder hot isostatic pressing (HIPping). The manufactured shield is a full-scale module No. 11a. Starting from a forged block of 1350 mm x 1300 mm x 450 mm, the main machining steps as deep drilling (1200 mm), 3D machining and sawing were performed. Tubes were 3D bent and large number of small parts were designed and machined. By welding together all the sub-parts we erected the main part of the water coolant circuit. Once the water circuit was built; the shield was completed using powder HIPping together with forged block embedding the tubes in a final solid part. The powder/solid HIP is used to minimize the number of BM seal welds in front of plasma. It increases the reliability of the components during operation. About 300 kg of stainless steel powder was densified together with the forged block. 3D measurement was done before and after the HIP cycle to collect the data to be compared with theoretical model. It allows to predict the main distortions of the solid bulk. Ultrasonic examination of the densified powder on the stainless steel bulk and around the bended tubes was performed as well as mechanical characterization of the samples. The recess for stub key attachment on the vacuum vessel side, the hydraulic connector, the key for the primary wall panel attachment on the front side and the link between the four parallel water coolant circuits were then machined to achieve the shield prototype

  12. 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)

    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)

  13. Preparatory works for PFBR shielding experiments Phase-IV at Apsara

    International Nuclear Information System (INIS)

    Proto-type Fast Breeder Reactor houses Radial and Axial shields inside reactor vessel to reduce the neutron flux impingement on in-vessel intermediate heat exdtanger (IHX)and toredure the activation of sodium in seoondarysystern. PFBR bulks have bcen calTied out in the shielding corner of Apsara reactor for optimizing the in-vessel radial and axial shielding. 10 experiments were conducted using shielding models of various combinations of steel, sodium, graphite and boron carbide to simulate the in-vessel radial and axial shielding in PFBR. These experiments have provided valuable data for design, in the form of bias factors to be used in the shield design. The extrapolate the bias factor to any further design change or for the design of further FBR 500 and larger size 1000 MWe fast reactors, it is essential to study the neutron transport through single shield materials. Towards this end, 10 more experiments are being conducted with single material shield models consisting of Cs, Cast Iron, graphite, B4C, Borated Graphite, Na, SS, Ni and Cr. An experiment simulating the actual shield geomctry in PFBR sub-assembly is also planned. As done in case of earlier experiments, detailed safety review and a comprehensive preparatory work were carried out prior to the commencement of the ongoing experiments as well, in order to ensure the safe and timely completion of the same. Based on experience gained in the earlier experiments certain design innovations and safety features the Phase-IV experiments with single shield models. (author)

  14. A very personal view of the development of radiation shielding theory

    International Nuclear Information System (INIS)

    Most research in radiation shielding has been project driven, aimed at providing answers important to the success of a specific type of facility or device. From the beginning, however, some attention has also been paid to fundamental shielding research - the study of how the radiation of interest penetrates through substantial thicknesses of matter. It is intended here to outline the history of theoretical shielding research of this nature, especially those developments in which the author has been either a participant or an interested spectator

  15. Bulk chemicals from biomass

    NARCIS (Netherlands)

    Haveren, van J.; Scott, E.L.; Sanders, J.P.M.

    2008-01-01

    Given the current robust forces driving sustainable production, and available biomass conversion technologies, biomass-based routes are expected to make a significant impact on the production of bulk chemicals within 10 years, and a huge impact within 20-30 years. In the Port of Rotterdam there is a

  16. Ferromagnetic bulk glassy alloys

    International Nuclear Information System (INIS)

    This paper deals with the review on the formation, thermal stability and magnetic properties of the Fe-based bulk glassy alloys in as-cast bulk and melt-spun ribbon forms. A large supercooled liquid region over 50 K before crystallization was obtained in Fe-(Al, Ga)-(P, C, B, Si), Fe-(Cr, Mo, Nb)-(Al, Ga)-(P, C, B) and (Fe, Co, Ni)-Zr-M-B (M=Ti, Hf, V, Nb, Ta, Cr, Mo and W) systems and bulk glassy alloys were produced in a thickness range below 2 mm for the Fe-(Al, Ga)-(P, C, B, Si) system and 6 mm for the Fe-Co-(Zr, Nb, Ta)-(Mo, W)-B system by copper-mold casting. The ring-shaped glassy Fe-(Al, Ga)-(P, C, B, Si) alloys exhibit much better soft magnetic properties as compared with the ring-shaped alloy made from the melt-spun ribbon because of the formation of the unique domain structure. The good combination of high glass-forming ability and good soft magnetic properties indicates the possibility of future development as a new bulk glassy magnetic material

  17. Heat transport in bulk/nanoporous/bulk silicon devices

    Energy Technology Data Exchange (ETDEWEB)

    Criado-Sancho, M. [Departamento de Ciencias y Técnicas Físicoquimicas, Facultad de Ciencias, UNED, Senda del Rey 9, 20040 Madrid (Spain); Jou, D., E-mail: David.Jou@uab.cat [Departament de Física, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Institut d' Estudis Catalans, Carme 47, 08001 Barcelona, Catalonia (Spain)

    2013-02-04

    We study heat transport in bulk/nanoporous/bulk silicon devices; we show that, despite bulk/nanoporous devices may act as thermal rectifiers, the non-linear aspects of their joint thermal conductance are not strong enough to lead to a negative differential thermal resistance, necessary to allow bulk/nanoporous/bulk Si devices to act as thermal transistors. Furthermore, we explicitly study the effective thermal conductivity of the mentioned devices for several temperatures, geometries, porosities, and pore size.

  18. 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.

  19. Neutron production, shielding and activation

    International Nuclear Information System (INIS)

    This chapter contains information on neutron cross-sections, production, spectra and yields; detection and detectors; shielding with various materials, particularly with ordinary concrete; and neutron activation products of interest to health physicists. Neutron energy terminology as well as neutron energy spectrum calculations are included

  20. Validity assessment of shielding design tools for ITER through analysis of benchmark experiment on SS316/water shield conducted at FNS/JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Verzilov, Y.M.; Konno, Chikara; Wada, Masayuki; Maekawa, Hiroshi; Oyama, Yukio; Uno, Yoshitomo [Japan Atomic Energy Research Inst., Ibaraki (Japan)

    1996-12-31

    To assess validity of the shielding design tools for ITER, the benchmark experiment on SS316/water shield conducted at FNS/JAERI is analyzed. As far as a simple bulk shield of SS316/water is concerned, the followings are found assuming that no uncertainty is involved in the response functions of the design parameters. Nuclear data bases of JENDL Fusion File and FENDL/E-1.0 are valid to predict all the design parameters with uncertainties less than a factor of 1.25. At the connection legs between shield blanket modules and back plates, both MCNP and DOT calculations can predict helium production rate with uncertainties less than 10%. For the toroidal field coils on the midplane, all the nuclear parameters can be predicted with uncertainties less than a factor of 1.25 by MCNP and DOT with consideration of self-shielding correction of cross sections and energy group structure of 125-n and 40-{gamma}. The uncertainties for toroidal field coils are considerably smaller than the design margins secured to the shielding designs under ITER/EDA. 22 refs., 8 figs.

  1. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Different theoretical and technological concepts and problems in using glass as a shielding material was discussed, some primarily designs for different types of radiation shielding windows were illustrated. (author)

  2. A Novel Radiation Shielding Material Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Radiation shielding simulations showed that epoxy loaded with 10-70% polyethylene would be an excellent shielding material against GCRs and SEPs. Milling produced...

  3. Fast Neutron Tomography of Low-Z Object in High-Z Material Shielding

    Science.gov (United States)

    Babai, Ruth Weiss; Sabo-Napadensky, Iris; Bar, Doron; Mor, Ilan; Tamim, Noam; Dangendorf, Volker; Tittelmeier, Kai; Bromberger, Benjamin; Weierganz, Mathias

    The technique and first results of Fast Neutron Tomography (FNCT) experiments are presented which are performed at the accelerator facility of PTB, Germany. A high-intensity neutron beam of broad spectral distribution with an average energy of 5.5 MeV, was produced by 11.5 MeV deuterons impinging upon a thick beryllium target. The capability of FNCT for high contrast imaging of low-Z materials embedded in thick high-Z shielding materials is demonstrated, which is superior to more conventional high-energy X-ray imaging techniques. For demonstrating the method special test objects were prepared: One consisted of an assembled polyethylene cylinder with holes of various diameters and directions drilled in its surface and inner parts. The plastic phantom was inserted into lead cylinders of different thicknesses. The detector system consisted of a plastic scintillator along with a dedicated optics, image-intensifier and a CCD camera. Two scintillator screens were compared: a bulk plastic scintillator screen and a fibres optical scintillator screen. The tomographic scans were taken in two geometrical configurations: cone beam and semi-fan beam configuration. The image quality favours the semi-fan beam configuration which on the other hand is more time consuming The obtained tomographic images and a comparison of the imaging quality between the different experimental conditions will be presented.

  4. National Low-Temperature Neutron Irradiation Facility (NLTNIF): status of development

    International Nuclear Information System (INIS)

    In May 1983 the Division of Materials Sciences, Office of Basic Energy Sciences of the Department of Energy authorized the establishment of a National Low-Temperature Neutron Irradiation Facility (NLTNIF) at ORNL's Bulk Shielding Reactor (BSR). The NLTNIF, which will be available for qualified experiments at no cost to users, will provide a combination of high radiation intensities and special environmental and testing conditions that have not been previously available in the US. Since the DOE authorization, work has proceeded on the design and construction of the new facility without interruption. This report describes the present status of the development of NLTNIF and, for the information of new candidate users, a recounting of the major specifications and capabilities is also given

  5. National Low-Temperature Neutron Irradiation Facility (NLTNIF). The status of development

    International Nuclear Information System (INIS)

    In May 1983, the Department of Energy authorized the establishment of a National Low-Temperature Neutron Irradiation Facility (NLTNIF) at ORNL's Bulk Shielding Reactor (BSR). The NLTNIF, which will be available for qualified experiments at no cost to users, will provide a combination of high radiation intensities and special environmental and testing conditions that have not been previously available in the US. Since the DOE authorization, work has proceeded on the design and construction of the new facility without interruption. This report describes the present status of the development of the NLTNIF and the anticipated schedule for completion and performance testing. There is a table of the major specifications and capabilities and a schematic layout of the irradiation cryostate for design and dimensioning of test and experiment assemblies

  6. The study on mix radio design and construction technology of radiation-shielding and high-density concrete

    International Nuclear Information System (INIS)

    Newly-constructed nuclear facilities requires the shielding concrete with density of 4600 kg/m3 or even higher for shielding of γ rays or neutron rays. Systemic tests and studies on radiation shielding concrete (neutrons and γ-ray absorbing) were conducted in such aspects as mix ratio design, preparation, construction technology, shielding effect, uniform shielding etc. The results show concrete for γ ray could be prepared with an average density of 4670 kg/m3, compressive strength of 37 MPa and permeability-resistant grade of P10. For neutron ray shield, the prepared concrete could be at an average density of 4680 kg/m3, with crystal water of 2.65% (wt) and boron of 0.11% (wt), and compressive strength of 45.6 MPa. (authors)

  7. Bulk materials handling review

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-02-15

    The paper provides details of some of the most important coal handling projects and technologies worldwide. It describes development by Aubema Crushing Technology GmbH, Bedeschi, Cimbria Moduflex, DBT, Dynamic Air Conveying Systems, E & F Services, InBulk Technologies, Nord-Sen Metal Industries Ltd., Pebco Inc, Primasonics International Ltd., R.J.S. Silo Clean (International) Ltd., Takraf GmbH, and The ACT Group. 17 photos.

  8. Reactor cavity cleanup system shielded filter installation

    International Nuclear Information System (INIS)

    The Seabrook Station reactor cavity cleanup system provides a flow path for refueling pool purification and drain down during plant refueling evolutions. The original system design included refueling pool surface skimmers and drains, a skimmer pump, an unshielded duplex basket type pump suction strainer and interconnecting stainless steel piping. The piping design utilized socket welded joints in small bore pipe with diaphragm values installed in the horizontal pipe runs downstream of the skimmer pump. The previously installed unshielded strainer in addition to the skimmer pump downstream piping components were determined to be inconsistent with Seabrook's proactive approach to dose reduction. To be consistent with ALARA (As Low As Reasonably Achievable) policy, a plant design change was authorized to install a lead shielded filter unit as a replacement for the existing duplex strainer. This filter unit, which utilizes multiple micron rating disposable basket type cartridges, has a threefold function of protecting the skimmer pump from large solids, providing bulk filtration of activated corrosion products from the refueling water in order to minimize CRUD buildup in downstream components, and enabling retrieval of foreign material drawn into the refueling pool drains

  9. Isotopic signatures by bulk analyses

    International Nuclear Information System (INIS)

    Los Alamos National Laboratory has developed a series of measurement techniques for identification of nuclear signatures by analyzing bulk samples. Two specific applications for isotopic fingerprinting to identify the origin of anthropogenic radioactivity in bulk samples are presented. The first example is the analyses of environmental samples collected in the US Arctic to determine the impact of dumping of radionuclides in this polar region. Analyses of sediment and biota samples indicate that for the areas sampled the anthropogenic radionuclide content of sediments was predominantly the result of the deposition of global fallout. The anthropogenic radionuclide concentrations in fish, birds and mammals were very low. It can be surmised that marine food chains are presently not significantly affected. The second example is isotopic fingerprinting of water and sediment samples from the Rocky Flats Facility (RFP). The largest source of anthropogenic radioactivity presently affecting surface-waters at RFP is the sediments that are currently residing in the holding ponds. One gram of sediment from a holding pond contains approximately 50 times more plutonium than 1 liter of water from the pond. Essentially 100% of the uranium in Ponds A-1 and A-2 originated as depleted uranium. The largest source of radioactivity in the terminal Ponds A-4, B-5 and C-2 was naturally occurring uranium and its decay product radium. The uranium concentrations in the waters collected from the terminal ponds contained 0.05% or less of the interim standard calculated derived concentration guide for uranium in waters available to the public. All of the radioactivity observed in soil, sediment and water samples collected at RFP was naturally occurring, the result of processes at RFP or the result of global fallout. No extraneous anthropogenic alpha, beta or gamma activities were detected. The plutonium concentrations in Pond C-2 appear to vary seasonally

  10. Overview of the SHIELDS Project at LANL

    Science.gov (United States)

    Jordanova, V.; Delzanno, G. L.; Henderson, M. G.; Godinez, H. C.; Jeffery, C. A.; Lawrence, E. C.; Meierbachtol, C.; Moulton, D.; Vernon, L.; Woodroffe, J. R.; Toth, G.; Welling, D. T.; Yu, Y.; Birn, J.; Thomsen, M. F.; Borovsky, J.; Denton, M.; Albert, J.; Horne, R. B.; Lemon, C. L.; Markidis, S.; Young, S. L.

    2015-12-01

    The near-Earth space environment is a highly dynamic and coupled system through a complex set of physical processes over a large range of scales, which responds nonlinearly to driving by the time-varying solar wind. Predicting variations in this environment that can affect technologies in space and on Earth, i.e. "space weather", remains a big space physics challenge. We present a recently funded project through the Los Alamos National Laboratory (LANL) Directed Research and Development (LDRD) program that is 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 specify the dynamics of the hot (keV) particles (the seed population for the radiation belts) on both macro- and micro-scale, including important physics of rapid particle injection and acceleration associated with magnetospheric storms/substorms and plasma waves. This challenging problem is 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. New data assimilation techniques employing data from LANL instruments on the Van Allen Probes and geosynchronous satellites are developed in addition to physics-based models. This research will provide a framework for understanding of key radiation belt drivers that may accelerate particles to relativistic energies and lead to spacecraft damage and failure. The ability to reliably distinguish between various modes of failure is critically important in anomaly resolution and forensics. SHIELDS will enhance our capability to accurately specify and predict the near-Earth space environment where operational satellites reside.

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

    International Nuclear Information System (INIS)

    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)

  12. Handbook of radiation shielding data

    International Nuclear Information System (INIS)

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

  13. Light shield for solar concentrators

    Energy Technology Data Exchange (ETDEWEB)

    Plesniak, Adam P.; Martins, Guy L.

    2014-08-26

    A solar receiver unit including a housing defining a recess, a cell assembly received in the recess, the cell assembly including a solar cell, and a light shield received in the recess and including a body and at least two tabs, the body defining a window therein, the tabs extending outward from the body and being engaged with the recess, wherein the window is aligned with the solar cell.

  14. How Concentration Shields Against Distraction

    OpenAIRE

    Sörqvist, Patrik; Marsh, John E.

    2015-01-01

    In this article, we outline our view of how concentration shields against distraction. We argue that higher levels of concentration make people less susceptible to distraction for two reasons. One reason is that the undesired processing of the background environment is reduced. For example, when people play a difficult video game, as opposed to an easy game, they are less likely to notice what people in the background are saying. The other reason is that the locus of attention becomes more st...

  15. Paramagnetism shielding in drilling fluid

    OpenAIRE

    Li, Zhuo

    2013-01-01

    In drilling operations, drilling fluid containing magnetic materials is used when drilling a well. The materials can significantly shield the Earth’s magnetic field as measured by magnetic sensors inside the drilling strings. The magnetic property of the drilling fluid is one of the substantial error sources for the determination of magnetic azimuth for wellbores. Both the weight material, cuttings, clay and other formation material plus metal filings from the tubular wear m...

  16. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  17. ATLAS Award for Shield Supplier

    CERN Multimedia

    2004-01-01

    ATLAS technical coordinator Dr. Marzio Nessi presents the ATLAS supplier award to Vojtech Novotny, Director General of Skoda Hute.On 3 November, the ATLAS experiment honoured one of its suppliers, Skoda Hute s.r.o., of Plzen, Czech Republic, for their work on the detector's forward shielding elements. These huge and very massive cylinders surround the beampipe at either end of the detector to block stray particles from interfering with the ATLAS's muon chambers. For the shields, Skoda Hute produced 10 cast iron pieces with a total weight of 780 tonnes at a cost of 1.4 million CHF. Although there are many iron foundries in the CERN member states, there are only a limited number that can produce castings of the necessary size: the large pieces range in weight from 59 to 89 tonnes and are up to 1.5 metres thick.The forward shielding was designed by the ATLAS Technical Coordination in close collaboration with the ATLAS groups from the Czech Technical University and Charles University in Prague. The Czech groups a...

  18. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

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

  19. Photonic Bandgap (PBG) Shielding Technology

    Science.gov (United States)

    Bastin, Gary L.

    2007-01-01

    Photonic Bandgap (PBG) shielding technology is a new approach to designing electromagnetic shielding materials for mitigating Electromagnetic Interference (EM!) with small, light-weight shielding materials. It focuses on ground planes of printed wiring boards (PWBs), rather than on components. Modem PSG materials also are emerging based on planar materials, in place of earlier, bulkier, 3-dimensional PBG structures. Planar PBG designs especially show great promise in mitigating and suppressing EMI and crosstalk for aerospace designs, such as needed for NASA's Constellation Program, for returning humans to the moon and for use by our first human visitors traveling to and from Mars. Photonic Bandgap (PBG) materials are also known as artificial dielectrics, meta-materials, and photonic crystals. General PBG materials are fundamentally periodic slow-wave structures in I, 2, or 3 dimensions. By adjusting the choice of structure periodicities in terms of size and recurring structure spacings, multiple scatterings of surface waves can be created that act as a forbidden energy gap (i.e., a range of frequencies) over which nominally-conductive metallic conductors cease to be a conductor and become dielectrics. Equivalently, PBG materials can be regarded as giving rise to forbidden energy gaps in metals without chemical doping, analogous to electron bandgap properties that previously gave rise to the modem semiconductor industry 60 years ago. Electromagnetic waves cannot propagate over bandgap regions that are created with PBG materials, that is, over frequencies for which a bandgap is artificially created through introducing periodic defects

  20. Steam generator hand hole shielding.

    Science.gov (United States)

    Cox, W E

    2000-05-01

    Seabrook Station is an 1198 MWE Pressurized Water Reactor (PWR) that began commercial operation in 1990. Expensive and dose intensive Steam Generator Replacement Projects among PWR operators have led to an increase in steam generator preventative maintenance. Most of this preventative maintenance is performed through access ports in the shell of the steam generator just above the tube sheet known as secondary side hand holes. Secondary side work activities performed through the hand holes are typically performed without the shielding benefit of water in the secondary side of the steam generator. An increase in cleaning and inspection work scope has led to an increase in dose attributed to steam generator secondary side maintenance. This increased work scope and the station goal of maintaining personnel radiation dose ALARA led to the development of the shielding concept described in this article. This shield design saved an estimated 2.5 person-rem (25 person-Smv) the first time it was deployed and is expected to save an additional 50 person-rem (500 person-mSv) over the remaining life of the plant. PMID:10770158

  1. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  2. Seal device for shield plug

    International Nuclear Information System (INIS)

    Purpose: To surely seal cover gases at a position nearer to the reactor core of a shield plug in LMFBR type reactors. Constitution: A shield plug is formed with through holes for insertion of a stopper or a through-cylinder. A step is provided to the through hole at the interium of the thickness of the shield plug and a seal ring is disposed on the step. The seal ring is retained on the side of the stopper or the through-cylinder by means of a holding member. The seal ring is urged to the step of the stopper by the own weight of the stopper or the through-cylinder to thereby seal the cover gases. Since the seal ring is retained on the side of the stopper or the through-cylinder, the seal ring is pulled up together with the extraction of the stopper or the through-cylinder and can be maintained or repaired with ease. (Ikeda, J.)

  3. RF-transparent solar shield

    International Nuclear Information System (INIS)

    By combining durable Kapton films with quartz fibers, an effective solar shield or blanket is produced which also serves as an efficient RF-transparent window. The window consists of a series of Kapton film envelopes sandwiching thin quartz paper. Not only must the window prevent the sun from overheating the electronics and distorting mechanically aligned antennas, it must also prevent radiant heat loss from inside the satellite when it is in shadow and radiating to space at approx. 40K. The guidelines for achieving an effective high-frequency RF window are a low dielectric constant to keep reflections down, a low loss tangent so RF absorption and molecular movement will be minimal, and low mass with tin and lightweight materials. Because these guidelines were followed, the RF insertion loss of the multiple envelope shield is less than 1/4 dB at high frequency. This paper concentrates on the material and processing aspects of an RF-transparent solar shield

  4. Facile Formation of High-Quality InGaN/GaN Quantum-Disks-in-Nanowires on Bulk-Metal Substrates for High-Power Light-Emitters.

    Science.gov (United States)

    Zhao, Chao; Ng, Tien Khee; Wei, Nini; Prabaswara, Aditya; Alias, Mohd S; Janjua, Bilal; Shen, Chao; Ooi, Boon S

    2016-02-10

    High-quality nitride materials grown on scalable and low-cost metallic substrates are considerably attractive for high-power light-emitters. We demonstrate here, for the first time, the high-power red (705 nm) InGaN/GaN quantum-disks (Qdisks)-in-nanowire light-emitting diodes (LEDs) self-assembled directly on metal-substrates. The LEDs exhibited a low turn-on voltage of ∼2 V without efficiency droop up to injection current of 500 mA (1.6 kA/cm(2)) at ∼5 V. This is achieved through the direct growth and optimization of high-quality nanowires on titanium (Ti) coated bulk polycrystalline-molybdenum (Mo) substrates. We performed extensive studies on the growth mechanisms, obtained high-crystal-quality nanowires, and confirmed the epitaxial relationship between the cubic titanium nitride (TiN) transition layer and the hexagonal nanowires. The growth of nanowires on all-metal stack of TiN/Ti/Mo enables simultaneous implementation of n-metal contact, reflector, and heat sink, which greatly simplifies the fabrication process of high-power light-emitters. Our work ushers in a practical platform for high-power nanowires light-emitters, providing versatile solutions for multiple cross-disciplinary applications that are greatly enhanced by leveraging on the chemical stability of nitride materials, large specific surface of nanowires, chemical lift-off ready layer structures, and reusable Mo substrates. PMID:26745217

  5. Facile Formation of High-quality InGaN/GaN Quantum-disks-in-Nanowires on Bulk-Metal Substrates for High-power Light-emitters

    KAUST Repository

    Zhao, Chao

    2016-01-08

    High-quality nitride materials grown on scalable and low-cost metallic substrates are considerably attractive for high-power light emitters. We demonstrate here, for the first time, the high-power red (705 nm) InGaN/GaN quantum-disks (Qdisks)-in-nanowire light-emitting diodes (LEDs) self-assembled directly on metal-substrate. The LEDs exhibited a low turn-on voltage of ~2 V without efficiency droop up to injection current of 500 mA (1.6 kA/cm2) at ~5 V. This is achieved through the direct growth and optimization of high-quality nanowires on titanium (Ti) coated bulk polycrystalline-molybdenum (Mo) substrates. We performed extensive studies on the growth mechanisms, obtained high-crystal-quality nanowires, and confirmed the epitaxial relationship between the cubic titanium nitride (TiN) transition layer and the hexagonal nanowires. The growth of nanowires on all-metal stack of TiN/Ti/Mo enables simultaneous implementation of n-metal contact, reflector and heat-sink, which greatly simplifies the fabrication process of high-power light emitters. Our work ushers in a practical platform for high-power nanowires light emitters, providing versatile solutions for multiple cross-disciplinary applications that are greatly enhanced by leveraging on the chemical stability of nitride materials, large specific surface of nanowires, chemical lift-off ready layer structures, and reusable Mo substrates.

  6. Shielding calculations for a production target for secondary beams

    Energy Technology Data Exchange (ETDEWEB)

    Rehm, K.E.; Back, B.B.; Jiang, C.L. [and others

    1995-08-01

    In order to estimate the amount of shielding required for a radioactive beam facility dose rate were performed. The calculations for production targets with different geometries were performed. The calculations were performed with the MSU shielding code assuming a 500-p{mu}A 200-MeV deuteron beam stopped in a thick Al target. The target and the ion-optical elements for beam extraction are located in a 2 m{sup 3} large volume at the center of the production cell. These dose rate calculations show that with a combination of Fe and concrete it is possible to reduce the dose rate expected at the surface of a 7-m-wide cube housing the production target to less than 2 mrem/hr.

  7. Verification of the shielding built for a Cyclotron accelerator

    International Nuclear Information System (INIS)

    According to the National Nuclear Energy Commission (CNEN) resolution 112/2011, administrative controls must be applied during the construction of a cyclotron and documents must be created showing that the facility can operate without radiological risks, referring even to the shielding efficiency. This study aimed to perform the analysis of the construction and efficiency of the bunker built for shielding, in the cyclotron of University of Sao Paulo Medical School Health System. This was possible through the measurements of a radiometric survey in normal working conditions, and testing related to compression resistance and density. The results showed that the compression resistance of the concrete used is higher than the expected value and the average density value obtained is within the tolerated limits. The radiometric survey results showed that the levels of ionizing radiation are well below the established limits. (author)

  8. Field maintenance of radiation-shielding windows at HFEF

    International Nuclear Information System (INIS)

    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

  9. Cost optimization of radiation processing facility - a study

    International Nuclear Information System (INIS)

    The radiation processing facilities after successful technology demonstration are undergoing through the transition phase of commercialisation. Financial viability of product irradiation is one of the important aspect to be looked into for the same. The cost of project needs to be optimised so as to make the scheme commercially viable. A case study with three alternates for proposed facility for agricultural products at RRCAT is presented in the paper. Construction plans with different layouts and shielding schemes were worked out. It is found that there is a saving of 35% of cost of construction in under ground composite shielding structure as compared to that of over ground shielding structure for this facility . (author)

  10. Shielding considerations for the 750-MeV electron accelerator at the University of Illinois

    International Nuclear Information System (INIS)

    This report summarizes some of the calculations that were carried out to provide shielding data for the 750-MeV electron accelerator under construction at the University of Illinois. All of the results described herein were obtained for a 300-MeV and/or 750-MeV electron beam. All calculations deal with doses produced by the particle beam during operation and do not include secondary radiation sources, i.e., induced radioactivity. The dose equivalents were obtained as a function of shield thickness so that various accident scenatios could be considered, i.e., various percentages of beam loss during operation. The calculated results that were considered included: (1) the earth shielding thickness (and iron door) surrounding the accelerator vault, (2) the earth shielding thickness around the beam transport tunnel, (3) an estimate of the thickness and composition of the movable shielding door in the general purpose electron beam experimental area, (4) the shield thickness around the beam dump in the bremsstrahlung irradiation facility, (5) skyshine dose from some of the experimental areas, and (6) dose rates inside and outside the tagged photon facility. The programs and cross section data bases used in the calculations, as well as the source neutron spectra calculations, are presented. The results of the dose calculations are presented and discussed

  11. Proposal for a national underground science facility

    International Nuclear Information System (INIS)

    The idea is explored of establishing a laboratory complex shielded from the cosmic ray flux at the earth's surface for the purpose of housing and providing technical support for experiments in particle physics, astrophysics, and other scientific disciplines. The scientific motivation for such an underground science facility is described, and the questions of location and desired properties of the facility are discussed

  12. 33 CFR 154.120 - Facility examinations.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Facility examinations. 154.120...) POLLUTION FACILITIES TRANSFERRING OIL OR HAZARDOUS MATERIAL IN BULK General § 154.120 Facility examinations. (a) The facility operator shall allow the Coast Guard, at any time, to make any examination and...

  13. Initial operational experience after installation of the DIII-D radiation shield

    International Nuclear Information System (INIS)

    The D3-D tokamak now operates with a neutron radiation shield to allow enhanced plasma operations with increased neutron production while minimizing the site boundary dose level. Neutron rates as high as 3 x 1015 neutrons/sec and total neutron production of 4 x 1015 neutrons per shot are obtained while maintaining the site dose below the DOE administrative level of 20 millirem per year; a much more restrictive level than the State of California radiation limits. The D3-D tokamak facility completed an upgrade in the spring of 1989 which included additional radiation shielding to allow more extensive operation with neutron generating deuterium beams and deuterium plasmas. The previous shielding, designed primarily as a gamma shield, consisted of 63 cm thick concrete walls to a height 3.1 m above the vessel midplane (∼1.7 m above the top of the plasma vessel). The additional neutron shield extended the concrete walls to a shielded roof consisting of a movable section and a fixed section. These additional shield walls are 30 cm of polyboron material while the roof is 36 cm of a water based boronated gel

  14. Novel light-weight materials for shielding gamma ray

    International Nuclear Information System (INIS)

    A comparison of gamma ray attenuation effectiveness of bulk aluminum, close-cell composite metal foams and open-cell aluminum foam infiltrated with variety of second phase materials were investigated and reported in this study. Mass attenuation coefficients for six sets of samples with three different areal densities of 2, 5 and 10 g/cm2 were determined at photon energies of 0.060, 0.662, 1.173, and 1.332 MeV. Theoretical values were calculated using XCOM software package. A complete agreement was observed between experimental and theoretical results. It is observed that close-cell composite metal foams exhibit a better shielding capability compared to open-cell Al foam with fillers. It is also observed that close-cell composite metal foams offer superior shielding effectiveness compared to bulk aluminum for energies below 0.662 MeV, the mass attenuation coefficients of steel–steel composite metal foam and Al–steel composite metal foam were measured 400 and 300% higher than that of aluminum A356. This study indicates the potential of utilizing the light-weight composite metal foams as shielding material replacing current heavy materials used for attenuation of low energy gamma ray with additional advantages such as high energy absorption and excellent heat rejection capabilities. - Highlights: • Close-cell metal foams were processed by powder metallurgy and casting techniques. • Open-cell foams were infiltrated with light weight fillers: wax, polyethylene, water. • Each material with three areal densities was studied under four photon energies. • Steel–steel composite foam is 400% more effective than aluminum against 241Am. • Al–steel composite foam is 300% more effective than aluminum against 241Am

  15. Wormholes in Bulk Viscous Cosmology

    OpenAIRE

    Jamil, Mubasher

    2008-01-01

    We investigate the effects of the accretion of phantom energy with non-zero bulk viscosity onto a Morris-Thorne wormhole. We have found that if the bulk viscosity is large then the mass of wormhole increases rapidly as compared to small or zero bulk viscosity.

  16. Recommendations for a Static Cosmic Ray Shield for Enriched Germanium Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Aguayo Navarrete, Estanislao; Orrell, John L.; Ankney, Austin S.; Berguson, Timothy J.

    2011-09-21

    This document provides a detailed study of cost and materials that could be used to shield the detector material of the international Tonne-scale germanium neutrinoless double-beta decay experiment from hadronic particles from cosmic ray showers at the Earth's surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during storage; in particular, when the detector material is being worked on at the detector manufacturer's facility. This work considers two options for shielding the detector material from cosmic ray particles. One option is to use a pre-existing structure already located near the detector manufacturer, such as Canberra Industries in Meriden, Connecticut. The other option is to build a shield onsite at a detector manufacturer's site. This paper presents a cost and efficiency analysis of such construction.

  17. Heat-shield design for glovebox applications

    International Nuclear Information System (INIS)

    Heat shields can often be used in place of insulation materials as an effective means of insulating glovebox furnace vessels. If used properly, shields can accomplish two important objectives: thermal insulation of the vessel to maintain a desired process temperature and protection of the glovebox, equipment, and user. A heat-shield assembly can be described as an arrangement of thin, properly-spaced, metal sheets that reduce radiation heat transfer. The main problem encountered in the design of a heat shield assembly is choosing the number of shields. In determining the heat transfer characteristics of a heat-shield assembly, a number of factors must be taken into consideration. The glovebox or outside environment, material properties, geometry, and operating temperature all have varying effects on the expected results. A simple method, for planar-horizontal and cylindrical-vertical shields, allowing the approximation of the outermost shield temperature, the practical number of shields, and the net heat-transfer rate will be presented. Methods used in the fabrication of heat-shield assemblies will also be discussed

  18. EMI Shields made from intercalated graphite composites

    Science.gov (United States)

    Gaier, James R.; Terry, Jennifer

    1995-01-01

    Electromagnetic interference (EMI) shielding typically makes up about twenty percent of the mass of a spacecraft power system. Graphite fiber/polymer composites have significantly lower densities and higher strengths than aluminum, the present material of choice for EMI shields, but they lack the electrical conductivity that enables acceptable shielding effectiveness. Bromine intercalated pitch-based graphite/epoxy composites have conductivities fifty times higher than conventional structural graphite fibers. Calculations are presented which indicate that EMI shields made from such composites can have sufficient shielding at less than 20% of the mass of conventional aluminum shields. EMI shields provide many functions other than EMI shielding including physical protection, thermal management, and shielding from ionizing radiation. Intercalated graphite composites perform well in these areas also. Mechanically, they have much higher specific strength and modulus than aluminum. They also have shorter half thicknesses for x-rays and gamma radiation than aluminum. Thermally, they distribute infra-red radiation by absorbing and re-radiating it rather than concentrating it by reflection as aluminum does. The prospects for intercalated graphite fiber/polymer composites for EMI shielding are encouraging.

  19. The Economics of Bulk Water Transport in Southern California

    OpenAIRE

    Andrew Hodges; Kristiana Hansen; Donald McLeod

    2014-01-01

    Municipalities often face increasing demand for limited water supplies with few available alternative sources. Under some circumstances, bulk water transport may offer a viable alternative. This case study documents a hypothetical transfer between a water utility district in northern California and urban communities located on the coast of central and southern California. We compare bulk water transport costs to those of constructing a new desalination facility, which is the current plan of ...

  20. Creating bulk nanocrystalline metal.

    Energy Technology Data Exchange (ETDEWEB)

    Fredenburg, D. Anthony (Georgia Institute of Technology, Atlanta, GA); Saldana, Christopher J. (Purdue University, West Lafayette, IN); Gill, David D.; Hall, Aaron Christopher; Roemer, Timothy John (Ktech Corporation, Albuquerque, NM); Vogler, Tracy John; Yang, Pin

    2008-10-01

    Nanocrystalline and nanostructured materials offer unique microstructure-dependent properties that are superior to coarse-grained materials. These materials have been shown to have very high hardness, strength, and wear resistance. However, most current methods of producing nanostructured materials in weapons-relevant materials create powdered metal that must be consolidated into bulk form to be useful. Conventional consolidation methods are not appropriate due to the need to maintain the nanocrystalline structure. This research investigated new ways of creating nanocrystalline material, new methods of consolidating nanocrystalline material, and an analysis of these different methods of creation and consolidation to evaluate their applicability to mesoscale weapons applications where part features are often under 100 {micro}m wide and the material's microstructure must be very small to give homogeneous properties across the feature.

  1. Explosive bulk charge

    Science.gov (United States)

    Miller, Jacob Lee

    2015-04-21

    An explosive bulk charge, including: a first contact surface configured to be selectively disposed substantially adjacent to a structure or material; a second end surface configured to selectively receive a detonator; and a curvilinear side surface joining the first contact surface and the second end surface. The first contact surface, the second end surface, and the curvilinear side surface form a bi-truncated hemispherical structure. The first contact surface, the second end surface, and the curvilinear side surface are formed from an explosive material. Optionally, the first contact surface and the second end surface each have a substantially circular shape. Optionally, the first contact surface and the second end surface consist of planar structures that are aligned substantially parallel or slightly tilted with respect to one another. The curvilinear side surface has one of a smooth curved geometry, an elliptical geometry, and a parabolic geometry.

  2. The Incredible Bulk

    CERN Document Server

    Fukushima, Keita; Kumar, Jason; Sandick, Pearl; Yamamoto, Takahiro

    2014-01-01

    Recent experimental results from the LHC have placed strong constraints on the masses of colored superpartners. The MSSM parameter space is also constrained by the measurement of the Higgs boson mass, and the requirement that the relic density of lightest neutralinos be consistent with observations. Although large regions of the MSSM parameter space can be excluded by these combined bounds, leptophilic versions of the MSSM can survive these constraints. In this paper we consider a scenario in which the requirements of minimal flavor violation, vanishing $CP$-violation, and mass universality are relaxed, specifically focusing on scenarios with light sleptons. We find a large region of parameter space, analogous to the original bulk region, for which the lightest neutralino is a thermal relic with an abundance consistent with that of dark matter. We find that these leptophilic models are constrained by measurements of the magnetic and electric dipole moments of the electron and muon, and that these models have ...

  3. Bulk muscles, loose cables.

    Science.gov (United States)

    Liyanage, Chamari R D G; Kodali, Venkata

    2014-01-01

    The accessibility and usage of body building supplements is on the rise with stronger internet marketing strategies by the industry. The dangers posed by the ingredients in them are underestimated. A healthy young man came to the emergency room with palpitations and feeling unwell. Initial history and clinical examination were non-contributory to find the cause. ECG showed atrial fibrillation. A detailed history for any over the counter or herbal medicine use confirmed that he was taking supplements to bulk muscle. One of the components in these supplements is yohimbine; the onset of symptoms coincided with the ingestion of this product and the patient is symptom free after stopping it. This report highlights the dangers to the public of consuming over the counter products with unknown ingredients and the consequential detrimental impact on health. PMID:25326558

  4. NPP bulk equipment dismantling problems and experience

    International Nuclear Information System (INIS)

    NPP bulk equipment dismantling problems and experience are summarized. 'ECOMET-S' JSC is shown as one of the companies which are able to make NPPs industrial sites free from stored bulk equipment with its further utilization. 'ECOMET-S' JSC is the Russian Federation sole specialized metallic LLW (MLLW) treatment and utilization facility. Company's main objectives are waste predisposal volume reduction and treatment for the unrestricted release as a scrap. Leningrad NPP decommissioned main pumps and moisture separators/steam super heaters dismantling results are presented. Prospective fragmentation technologies (diamond and electro-erosive cutting) testing results are described. The electro-erosive cutting machine designed by 'ECOMET-S' JSC is presented. The fragmentation technologies implementation plans for nuclear industry are presented too. (author)

  5. Custom shielding blocks - a prospective study

    International Nuclear Information System (INIS)

    In delivering radiation to the cancer patients, we need to shield the critical organs which come in the way of the radiation treatment portal. To avoid critical organs getting more than the tolerable dose, we use shielding blocks, which restrict the radiation to the tumour volume. Various metals are being used as shielding materials. The transmission through the shielding block should be less than 5% and the thickness of the material needed to achieve this transmission is 5 HVL. Lead is commonly used for shielding. Another material called Cerrobend (Low Melting Alloy) is used to prepare custom shielding blocks. The standard lead blocks supplied is of 5 cm thickness and for Cerrobend (LMA), the equivalent thickness required is 7.5 cm. In this paper, a comparison between the standard lead and LMA is described

  6. Highly heat-removing radiation shielding material

    International Nuclear Information System (INIS)

    Highly heat-removing radiation shielding material is constituted with fine particles prepared by coating metals of high heat conductivity to fine particles comprising materials having excellent radiation shielding performance. Then, the fine particles applied with the coating are mixed and filled in a shielding container or applied with hot press into a layerous form and used as a shielding member. In view of the above, since the coated fine particles provide the shielding performance against radiation such as neutrons and gamma rays, and the coating metals provide the heat removing performance, they act as a shielding material having heat removing performance as a whole. The combination of the coated fine particles and the coating metals are selected depending on the respective conditions for use. With such a constitution, radioactive wastes involving a problem of heat generation can be transported or stored safely. (T.M.)

  7. Dynamic rotating-shield brachytherapy

    International Nuclear Information System (INIS)

    Purpose: To present dynamic rotating shield brachytherapy (D-RSBT), a novel form of high-dose-rate brachytherapy (HDR-BT) with electronic brachytherapy source, where the radiation shield is capable of changing emission angles during the radiation delivery process.Methods: A D-RSBT system uses two layers of independently rotating tungsten alloy shields, each with a 180° azimuthal emission angle. The D-RSBT planning is separated into two stages: anchor plan optimization and optimal sequencing. In the anchor plan optimization, anchor plans are generated by maximizing the D90 for the high-risk clinical-tumor-volume (HR-CTV) assuming a fixed azimuthal emission angle of 11.25°. In the optimal sequencing, treatment plans that most closely approximate the anchor plans under the delivery-time constraint will be efficiently computed. Treatment plans for five cervical cancer patients were generated for D-RSBT, single-shield RSBT (S-RSBT), and 192Ir-based intracavitary brachytherapy with supplementary interstitial brachytherapy (IS + ICBT) assuming five treatment fractions. External beam radiotherapy doses of 45 Gy in 25 fractions of 1.8 Gy each were accounted for. The high-risk clinical target volume (HR-CTV) doses were escalated such that the D2cc of the rectum, sigmoid colon, or bladder reached its tolerance equivalent dose in 2 Gy fractions (EQD2 with α/β= 3 Gy) of 75 Gy, 75 Gy, or 90 Gy, respectively.Results: For the patients considered, IS + ICBT had an average total dwell time of 5.7 minutes/fraction (min/fx) assuming a 10 Ci192Ir source, and the average HR-CTV D90 was 78.9 Gy. In order to match the HR-CTV D90 of IS + ICBT, D-RSBT required an average of 10.1 min/fx more delivery time, and S-RSBT required 6.7 min/fx more. If an additional 20 min/fx of delivery time is allowed beyond that of the IS + ICBT case, D-RSBT and S-RSBT increased the HR-CTV D90 above IS + ICBT by an average of 16.3 Gy and 9.1 Gy, respectively.Conclusions: For cervical cancer patients, D

  8. Dynamic rotating-shield brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Yunlong [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 (United States); Flynn, Ryan T.; Kim, Yusung [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Yang, Wenjun [Department of Medical Physics, University of Wisconsin-Madison, 1111 Highland Avenue, Madison, Wisconsin 53705 (United States); Wu, Xiaodong [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 and Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States)

    2013-12-15

    Purpose: To present dynamic rotating shield brachytherapy (D-RSBT), a novel form of high-dose-rate brachytherapy (HDR-BT) with electronic brachytherapy source, where the radiation shield is capable of changing emission angles during the radiation delivery process.Methods: A D-RSBT system uses two layers of independently rotating tungsten alloy shields, each with a 180° azimuthal emission angle. The D-RSBT planning is separated into two stages: anchor plan optimization and optimal sequencing. In the anchor plan optimization, anchor plans are generated by maximizing the D{sub 90} for the high-risk clinical-tumor-volume (HR-CTV) assuming a fixed azimuthal emission angle of 11.25°. In the optimal sequencing, treatment plans that most closely approximate the anchor plans under the delivery-time constraint will be efficiently computed. Treatment plans for five cervical cancer patients were generated for D-RSBT, single-shield RSBT (S-RSBT), and {sup 192}Ir-based intracavitary brachytherapy with supplementary interstitial brachytherapy (IS + ICBT) assuming five treatment fractions. External beam radiotherapy doses of 45 Gy in 25 fractions of 1.8 Gy each were accounted for. The high-risk clinical target volume (HR-CTV) doses were escalated such that the D{sub 2cc} of the rectum, sigmoid colon, or bladder reached its tolerance equivalent dose in 2 Gy fractions (EQD2 with α/β= 3 Gy) of 75 Gy, 75 Gy, or 90 Gy, respectively.Results: For the patients considered, IS + ICBT had an average total dwell time of 5.7 minutes/fraction (min/fx) assuming a 10 Ci{sup 192}Ir source, and the average HR-CTV D{sub 90} was 78.9 Gy. In order to match the HR-CTV D{sub 90} of IS + ICBT, D-RSBT required an average of 10.1 min/fx more delivery time, and S-RSBT required 6.7 min/fx more. If an additional 20 min/fx of delivery time is allowed beyond that of the IS + ICBT case, D-RSBT and S-RSBT increased the HR-CTV D{sub 90} above IS + ICBT by an average of 16.3 Gy and 9.1 Gy, respectively

  9. 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)

  10. Shielding of medically used proton accelerators

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  12. Radiation dose reduction by water shield

    International Nuclear Information System (INIS)

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

  13. Shielding for thermoacoustic tomography with RF excitation

    Science.gov (United States)

    Mitchell, M.; Becker, G.; Dey, P.; Generotzky, J.; Patch, S. K.

    2008-02-01

    Radiofrequency (RF) pulses used to generate thermoacoustic computerized tomography (TCT) signal couple directly into the pulser-receiver and oscilloscope, swamping true TCT signal. We use a standard RF enclosure housing both RF amplifier and object being imaged. This is similar to RF shielding of magnetic resonance imaging (MRI) suites and protects electronics outside from stray RF. Unlike MRI, TCT receivers are ultrasound transducers, which must also be shielded from RF. A transducer housing that simultaneously shields RF and permits acoustic transmission was developed specifically for TCT. We compare TCT signals measured with and without RF shielding.

  14. Shielding integral benchmark archive and database

    International Nuclear Information System (INIS)

    SINBAD (Shielding integral benchmark archive and database) is a new electronic database developed to store a variety of radiation shielding benchmark data so that users can easily and incorporate the data into their calculations. SINBAD is an excellent data source for users who require the quality assurance necessary in developing cross-section libraries or radiation transport codes. The future needs of the scientific community are best served by the electronic database format of SINBAD and its user-friendly interface, combined with its data accuracy and integrity. It has been designed to be able to include data from nuclear reactor shielding, fusion blankets and accelerator shielding experiments. (authors)

  15. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

    To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252Cf isotopic neutron source. During calculations a detailed model for the 252Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

  16. Measurement of exposure to personnel performing shielding evaluations

    Directory of Open Access Journals (Sweden)

    Leland Page

    2014-03-01

    Full Text Available Purpose: Determine the exposure to personnel accrued while performing shielding evaluations of diagnostic facilities.Methods: Shielding inspections were performed at three sites: a small animal imaging facility, an imaging center consisting of general x-ray, CT, and PET/CT rooms, and a cardiac catheterization lab. The small animal facility was evaluated using two 20 mCi Tc-99m uncollimated sources. The imaging center was evaluated using a 20 mCi source of F-18 for the PET/CT room and a collimated 25 mCi source of Tc-99m for the CT and x-ray rooms. The catheterization lab was evaluated using an uncollimated 35 mCi vial of Tc-99m. Electronic personal dosimeters (Rados RAD-60R, RADOS Technology Oy Turku, Finland were worn by each of the personnel involved in performing shielding inspections. The cumulative exposure reading from the dosimeter for each surveyor was recorded.Results: The surveyor who positioned the open source at the small animal facility had an exposure reading of 3.5 mR in 5.5 hours. Those who were outside the room surveying the barriers had readings of 0 and 0.1 mR in 5 hours. For the cath lab, the surveyor positioning the source had a reading of 0.4 mR in 30 minutes, while those outside had readings of 0 mR. At the imaging center all personnel surveying the PET/CT room with an open F-18 source had readings of 2.0, 1.1, and 2.9 mR in 4 hours. Surveys of the CT and general radiography rooms with a collimated Tc-99m source had readings of 1.2, 0.9, and 0.2mR in 5 hours.Conclusion: The surveyors who were in the room with the sources had the highest exposures. Using a higher energy source (F-18 also led to higher exposures. By using a collimated source, the time to measure each barrier for penetrations was increased, but the surveyors’ exposures were  lower.---------------------------------------Cite this article as: Page L, Dodge C, Li G, Wendt R. Measurement of exposure to personnel performing shielding evaluations. Int J Cancer

  17. Experimental and computational study of ordinary concrete shielding properties

    International Nuclear Information System (INIS)

    An analysis of shielding properties for concrete in common use was performed because they depend highly on specific properties of components, production technology and operating conditions which are difficult to control reliably. Fast neutron and photon transmission were measured on the facility which includes bare heavy water experimental reactor with external neutron converter. Calculations were performed by ANISN code. Results were obtained for fast neutron flux, neutron dose equivalent and photon dose transmission. Variation of neutron relaxation length with concrete slab thickness was determined. (author)

  18. Research of Bulk Erase Operation in Vertical Three-Dimensional Cell Array Architecture

    Science.gov (United States)

    Yang, Hyung-jun; Lee, Gae-hun; Kim, Kyeong-rok; Song, Yun-heub

    2013-04-01

    A bit-cost scalable (BiCS) NAND flash memory with a bulk erasing method is investigated in view of cell characteristics and uniformity. The proposed cell array has an additional electrode layer for a bulk erase operation in the middle of a vertical channel string cell. Here, under a bias condition of 20 V, a programming threshold voltage of 4.2 V at 1 ms and an erasing threshold voltage of Vth = -1.5 V at 10 ms are confirmed, which is acceptable for flash memories. Furthermore, the shielding transistor close to an erase electrode is also investigated, which gives better erase characteristics for the cells adjacent to the erase electrode. From this result, we expect that a bulk erasable-BiCS technology with a shielding transistor can be a candidate three-dimensional (3D) NAND flash memory.

  19. Crackle template based metallic mesh with highly homogeneous light transmission for high-performance transparent EMI shielding

    Science.gov (United States)

    Han, Yu; Lin, Jie; Liu, Yuxuan; Fu, Hao; Ma, Yuan; Jin, Peng; Tan, Jiubin

    2016-05-01

    Our daily electromagnetic environment is becoming increasingly complex with the rapid development of consumer electronics and wireless communication technologies, which in turn necessitates the development of electromagnetic interference (EMI) shielding, especially for transparent components. We engineered a transparent EMI shielding film with crack-template based metallic mesh (CT-MM) that shows highly homogeneous light transmission and strong microwave shielding efficacy. The CT-MM film is fabricated using a cost-effective lift-off method based on a crackle template. It achieves a shielding effectiveness of ~26 dB, optical transmittance of ~91% and negligible impact on optical imaging performance. Moreover, high–quality CT-MM film is demonstrated on a large–calibre spherical surface. These excellent properties of CT-MM film, together with its advantages of facile large-area fabrication and scalability in processing on multi-shaped substrates, make CT-MM a powerful technology for transparent EMI shielding in practical applications.

  20. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment machine design incorporates comprehensive planning for efficient and safe component maintenance. Three programmatic decisions have been made to insure the successful implementation of this objective. First, the tokamak incorporates radiation shielding to reduce activation of components and limit the dose rate to personnel working on the outside of the machine. This allows most of the ex-vessel equipment to be maintained through conventional ''hands-on'' procedures. Second, to the maximum extent possible, low activation materials will be used inside the shielding volume. This resulted in the selection of Titanium (Ti-6Al-4V) for the vacuum vessel and PFC structures. The third decision stipulated that the primary in-vessel components will be replaced or repaired via remote maintenance tools specifically provided for the task. The component designers have been given the responsibility of incorporating maintenance design and for proving the maintainability of the design concepts in full-scale mockup tests prior to the initiation of final fabrication. Remote maintenance of the TPX machine is facilitated by general purpose tools provided by a special purpose design team. Major tools will include an in-vessel transporter, a vessel transfer system and a large component transfer container. In addition, tools such as manipulators and remotely operable impact wrenches will be made available to the component designers by this group. Maintenance systems will also provide the necessary controls for this equipment

  1. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) machine design incorporates comprehensive planning for efficient and safe component maintenance. Three programmatic decisions have been made to insure the successful implementation of this objective. First, the tokamak incorporates radiation shielding to reduce activation of components and limit the dose rate to personnel working on the outside of the machine. This allows most of the ex-vessel equipment to be maintained through conventional open-quotes hands-onclose quotes procedures. Second, to the maximum extent possible, low activation materials will be used inside the shielding volume. This resulted in the selection of Titanium (Ti-6Al-4V) for the vacuum vessel and Plasma Facing Components (PFC) structures. The third decision stipulated that the primary in-vessel components will be replaced or repaired via remote maintenance tools specifically provided for the task. The component designers have been given the responsibility of incorporating maintenance design and for proving the maintainability of the design concepts in full-scale mockup tests prior to the initiation of final fabrication. Remote maintenance of the TPX machine is facilitated by general purpose tools provided by a special purpose design team. Major tools will include an in-vessel transporter, a vessel transfer system and a large component transfer container. In addition, tools such as manipulators and remotely operable impact wrenches will be made available to the component designers by this group. Maintenance systems will also provide the necessary controls for this equipment

  2. Substituent effects on nuclear shielding

    International Nuclear Information System (INIS)

    The important role of nuclear magnetic resonance (NMR) spectroscopy in chemistry arises largely from the consequences of nuclear shielding. The fact that nuclei in different electronic environments have different nuclear shieldings, and hence different chemical shifts, makes NMR a powerful probe of electronic structure. Empirical rules relating chemical shifts to substituent (σ) constants, electron densities, electronegativities, and a variety of other empirical parameters have proven of great benefit to problems of organic structural elucidation, and to fundamental studies of molecular electron distributions. This review focuses on one specific application in the latter category -the study of substituent electronic effects on chemical shifts. The aim is not to provide a compendium of substituent effects on chemical shifts to aid in structural assignments, but to show how fine detail relating to the distribution and polarization of electrons in molecules may be determined from chemical shift studies. Chapters are devoted to 1H, 11B, 13C, 15N, 17O, 19F, 31P, 33S, 77Se, 95Mo and 199Hg chemical shifts. (U.K.)

  3. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lead used for shielding is often surface contaminated with radionuclides and is therefore a Resource Conservation and Recovery Act (RCRA) D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Lab. decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 100 metric tons and likely to grow substantially because of planned decommissioning operations. This lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for decontaminating lead is removing the thin superficial layer of contamination with an abrasive medium under pressure. For lead, a mixture of alumina with water and air at about 280 kPa (40 psig) rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a sealed-off area. The slurry of abrasive and particles of lead falls through a floor grating and is collected in a pump. A pump sends the slurry mixture back to the spray gun, creating a continuous process

  4. Comparison of different shielding materials used at proton accelerators and cost-benefit analysis

    International Nuclear Information System (INIS)

    During last decades physicians' and physicist's experience and confidence in proton beam radiotherapy has grown significantly. Construction of a number of new proton therapy facilities is already underway, and several are in planning stages for the near future. Cost-effective shielding design of these facilities is important. We present comparative analysis of different shielding materials that are typically used at proton accelerators and in proton radiotherapy facilities. We have used Geant4 tool-kit for simulation of the passage of particles through matter. We have analyzed shielding properties of iron, borated concrete, a water tank with borated polyethylene walls, borated polyethylene, and borated fire retardant plywood. We have simulated 240 MeV protons incident on a thin copper target that generated radiation fields of primary protons as well as secondaries produced in the target incident on the shielding block. We found that iron is most effective per unit length. It may be the most cost-effective option if one considers using the so called SEG blocks in combination with 1 to 2 ft of concrete layer. The material of the SEG block consists primarily of iron from recycled government facility metals, and is slightly radioactive. The slight inherent radioactivity as well as low energy (< 847 keV) secondary neutrons from iron will be shielded by a thin concrete layer. We also find that borated fire retardant plywood can be a cost-effective alternative for borated polyethylene in many shielding applications where borated polyethylene sheets are used that are arguably not less fire hazardous. (authors)

  5. CYLSEC: A three dimensional shield evaluation code

    International Nuclear Information System (INIS)

    Existing point kernel gamma codes are either limited to simple geometry configurations or require rather cumbersome input. These codes also require the user to specify the mesh size used in integrating the kernel. This results in computational inefficiencies since it is difficult to establish criteria for choosing mesh size and because it is generally not possible to assure convergence without solving the problem more than once. The interactive program CYLSEC was recently developed to improve this situation. CYLSEC can be used to evaluate bulk or local shielding for radioactive components, to treat streaming problems and to calculate a variety of gamma ray response functions. It will accept three dimensional geometries that can be described in terms of orthogonal slabs, right cylinders and/or right parallelepipeds. While the problem geometry is specified in rectangular coordinates, the integration of the kernel is performed in spherical coordinates. This allows explicit integration over the radial variable, thus reducing the problem to a double integral. The integral mesh size varies and is internally determined such that a specified convergence criterion is met. CYLSEC is also designed to recognize and take advantage of any problem symmetry in order to maximize efficiency. Program input is through interactive routines that are self checking and permit the user to make corrections. A gamma ray data library is provided, however, alternate data may be specified if desired. Comparisons between CYLSEC and other point kernel codes (QAD, GRACE) show excellent agreement in results and demonstrate that CYLSEC requires significantly less CPU time. Comparisons with the discrete ordinates code ANISN also show good agreement. An additional attraction to CYLSEC is that it is suitable for conversion to mini or personal computers

  6. Applications and modelling of bulk HTSs in brushless ac machines

    International Nuclear Information System (INIS)

    The use of high temperature superconducting material in its bulk form for engineering applications is attractive due to the large power densities that can be achieved. In brushless electrical machines, there are essentially four properties that can be exploited; their hysteretic nature, their flux shielding properties, their ability to trap large flux densities and their ability to produce levitation. These properties translate to hysteresis machines, reluctance machines, trapped-field synchronous machines and linear motors respectively. Each one of these machines is addressed separately and computer simulations that reveal the current and field distributions within the machines are used to explain their operation. (author)

  7. Improved Electromagnetic Interference Shielding Properties of MWCNT–PMMA Composites Using Layered Structures

    Directory of Open Access Journals (Sweden)

    Saini P

    2009-01-01

    Full Text Available Abstract Electromagnetic interference (EMI shielding effectiveness (SE of multi-walled carbon nanotubes–polymethyl methacrylate (MWCNT–PMMA composites prepared by two different techniques was measured. EMI SE up to 40 dB in the frequency range 8.2–12.4 GHz (X-band was achieved by stacking seven layers of 0.3-mm thick MWCNT–PMMA composite films compared with 30 dB achieved by stacking two layers of 1.1-mm thick MWCNT–PMMA bulk composite. The characteristic EMI SE graphs of the composites and the mechanism of shielding have been discussed. SE in this frequency range is found to be dominated by absorption. The mechanical properties (tensile, flexural strength and modulus of the composites were found to be comparable or better than the pure polymer. The studies therefore show that the composite can be used as structurally strong EMI shielding material.

  8. Improved Electromagnetic Interference Shielding Properties of MWCNT-PMMA Composites Using Layered Structures.

    Science.gov (United States)

    Pande, Shailaja; Singh, Bp; Mathur, Rb; Dhami, Tl; Saini, P; Dhawan, Sk

    2009-01-01

    Electromagnetic interference (EMI) shielding effectiveness (SE) of multi-walled carbon nanotubes-polymethyl methacrylate (MWCNT-PMMA) composites prepared by two different techniques was measured. EMI SE up to 40 dB in the frequency range 8.2-12.4 GHz (X-band) was achieved by stacking seven layers of 0.3-mm thick MWCNT-PMMA composite films compared with 30 dB achieved by stacking two layers of 1.1-mm thick MWCNT-PMMA bulk composite. The characteristic EMI SE graphs of the composites and the mechanism of shielding have been discussed. SE in this frequency range is found to be dominated by absorption. The mechanical properties (tensile, flexural strength and modulus) of the composites were found to be comparable or better than the pure polymer. The studies therefore show that the composite can be used as structurally strong EMI shielding material. PMID:20596500

  9. Radiation shielding design calculation of gamma knife for therapy

    International Nuclear Information System (INIS)

    The author reports the method and results of radiation shielding calculation of the gamma knife for therapy which is composed of thirty 60Co sources each with 7.4 EBq, semi-spherical shield, lateral shielding cupboard and the shielding door. The shielding thicknesses of the back shield, the lateral shielding cupboard and the shielding door were calculated. The leakage radiation by test indicates that the shielding is sufficient safety for this Gamma knife and the Kerma rate of control calculated agrees with that by test

  10. 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.

  11. Energy loss caused by shielding effect of steel cage outside source tube

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The energy loss, produced by shielding effect of steel cage outside the source tube, is quite considerable.With PENELOPE software package, MC results have been obtained based on the simulation of different source conformations. The result illustrates that the naked source tubes can improve the utilization ratio of the cobalt facilities. It demonstrates the applied value of the naked source tube in engineering.

  12. Historical waste - biological shield and documentation during decommissioning - 59056

    International Nuclear Information System (INIS)

    Document available in abstract form only. Full text of publication follows: Inventory records of isotopes in radioactive waste are important as documentation with respect to governmental control, final disposal and public transparency. We present a simple, practical and cost-effective method for characterization of a part of the radioactive waste from decommissioning of a research reactor: The biological shielding. The method uses documentation from the decommissioning and from the construction drawings and blueprints of the reactor as well as measurements based on samples from the facility. The data presented is from a 5 MW experimental light water nuclear reactor (DR-2) shutdown in 1975 and decommissioned in 2008. The method incorporates an activity distribution in the biological shield. The distribution is based on measurements of samples from across the shielding and coupled with the distance to the center of the reactor core. The exact origin of each waste item is determined from pictures from the decommissioning and from old blueprints and construction drawings of the reactor. The uncertainty and usefulness of the method is related directly to different factors such as: The amount of samples obtained and the position of these with respect to the origin of the waste, the accuracy of the documentation of the decommissioning, the size of the waste items

  13. 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.)

  14. X-ray face mask and chest shield device

    International Nuclear Information System (INIS)

    A protective face mask is disclosed that is designed to shield an x-ray technician or machine operator primarily from random secondary or scatter x-rays deflected toward his face, head and neck by the table, walls, equipment and other reflecting elements in an x-ray room or chamber, during the period of exposure while adjacent the object or person being exposed to the x-ray beam. The face mask and chest shield device can be mounted on a patient's shoulders in reverse attitude to protect the back of a patient's head and neck from the x-ray beam while being exposed to such beam for chest or upper body portion study and examination. The face mask is relatively or substantially transparent and contains lead in combination with a plastic ionomer or comonomer, which absorbs or resists penetration, to a degree, of the random deflected secondary or scatter x-rays or the x-ray beam through the mask. The face mask is removably attachable to the chest shield for facile application of the device to and support upon the shoulders of the technician or the patient

  15. Diagnostic X-Ray shielding : new data and new concepts

    International Nuclear Information System (INIS)

    Report No. 49 of the National Council on Radiation Protection and Measurement (NCRP 49, 1976) is currently the primary source of information to be used for design of protective barrier shielding for diagnostic x-ray facilities in the USA. However, a number of more recent publications have presented new methodologies and attenuation data for alternative shielding materials not discussed in Report No. 49. A nonlinear analytical model of photon transmission that accurately fits and parameterizes attenuation data has been developed. Use of this model simplifies primary and secondary barrier calculations and permits 'exact' solutions for secondary barriers. This presentation will describe the use of this model and present a recently acquired database of single and three phase attenuation measurements obtained with state-of-the-art monitoring equipment. This data set, which contains measurements from six diagnostic x-ray shielding materials, will be compared with experimental and theoretical transmission data found in the literature. The model, which has been shown to provide a robust and accurate method of portraying experimental attenuation data, will be used in the revision of NCRP 49 to specify the transmission characteristics of all protective materials. (author)

  16. Inflation from bulk viscosity

    CERN Document Server

    Bamba, Kazuharu

    2015-01-01

    We explore the perfect fluid description of the inflationary universe. In particular, we investigate a fluid model with the bulk-viscosity term. We find that the three observables of inflationary cosmology: the spectral index of the curvature perturbations, the tensor-to-scalar ratio of the density perturbations, and the running of the spectral index, can be consistent with the recent Planck results. We also reconstruct the explicit equation of state (EoS) of the viscous fluid from the spectral index of the curvature perturbations compatible with the Planck analysis. In the reconstructed models of the viscous fluid, the tensor-to-scalar ratio of the density perturbations can satisfy the constraints obtained from the Planck satellite. The running of the spectral index can explain the Planck data. In addition, it is demonstrated that in the reconstructed models of the viscous fluid, the graceful exit from inflation can be realized. Furthermore, we show that the singular inflation can occur in the viscous fluid ...

  17. Bulk-Fill Resin Composites

    DEFF Research Database (Denmark)

    Benetti, Ana Raquel; Havndrup-Pedersen, Cæcilie; Honoré, Daniel;

    2015-01-01

    restorative procedure. The aim of this study, therefore, was to compare the depth of cure, polymerization contraction, and gap formation in bulk-fill resin composites with those of a conventional resin composite. To achieve this, the depth of cure was assessed in accordance with the International Organization...... for Standardization 4049 standard, and the polymerization contraction was determined using the bonded-disc method. The gap formation was measured at the dentin margin of Class II cavities. Five bulk-fill resin composites were investigated: two high-viscosity (Tetric EvoCeram Bulk Fill, SonicFill) and...... three low-viscosity (x-tra base, Venus Bulk Fill, SDR) materials. Compared with the conventional resin composite, the high-viscosity bulk-fill materials exhibited only a small increase (but significant for Tetric EvoCeram Bulk Fill) in depth of cure and polymerization contraction, whereas the low...

  18. Water confined in carbon nanotubes: Magnetic response and proton chemical shieldings

    Science.gov (United States)

    Huang, Patrick; Schwegler, Eric; Galli, Giulia

    2009-03-01

    Carbon nanotubes (CNT) provide a well-defined environment for the study of confined water, whose behavior can differ markedly from bulk water. The application of nuclear magnetic resonance (NMR) to probe the local water structure and dynamics in these cases is hindered by ambiguities in the interpretation of the NMR spectra. We employ linear response theory to evaluate the ^1H chemical shieldings of liquid water in semiconducting CNTs, where the electronic structure is derived from density functional theory with periodic boundary conditions. The shieldings are sampled from trajectories generated via first-principles molecular dynamics simulations at ambient conditions, for water in CNTs with diameters d=11 åand 14.9 å@. We find a large (˜-23 ppm) upfield shift relative to bulk liquid water, which is a consequence of strongly anisotropic magnetic fields induced in the CNT by the applied magnetic field.

  19. Decommissioning of the Risoe hot cell facility

    International Nuclear Information System (INIS)

    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)

  20. The use of linked shielding codes to substantiate the design of the top corner shielding of a CAGR

    International Nuclear Information System (INIS)

    This paper presents a summary of the detailed design substantiation performed for the current commercial AGR internal shielding around the top corner region of the reactor. The design is required to reduce shutdown activation dose rates at accessible positions inside the reactor pressure vessel to the order of 1 mSv/h, whilst at the same time providing adequate space outside the shielding for the remote operation of in-service inspection equipment. This design substantiation work serves as an example of the use of the latest UK shielding codes and demonstrates their flexibility, user-oriented linking capabilities and their practicality as design tools. The codes used involve a variety of the standard techniques for the solution of radiation transport problems: neutron diffusion (SCORMA), kernel/albedo neutron streaming (MULTISORD), Monte Carlo neutron transport (McBEND) and point kernel neutron and gamma ray line-of-sight integration (RANKERN). The linking facilities utilised in this particular application include the automatic transfer of Monte Carlo neutron collision data as secondary gamma-ray source terms into a gamma-ray point kernel integration calculation. This technique means that the inherently accurate, but normally uneconomic, Monte Carlo method can be employed as a design tool in complex limited attentuation situations, as part of an integrated calculational route for large attenuation design problems. (author)

  1. Shielding Desing for Storage of Radioactive Waste from TRR-1/M1 Pool Refurbishment

    International Nuclear Information System (INIS)

    Four sets of bolts and nuts were discarded as radioactive waste from the Thailand Research Reactor TRR-1/M1 pool during the pool refurbishment in 2011. The exposure rate of the waste is roughly 10 R/hr at the contact point and mainly comes from Co-60. The waste is currently stored at the Radioactive Waste Management facility in a generic shield container. This work has been initiated to obtain an optimum design for the shield container which should require less space with lower cost in order to store this waste long-term. The Monte Carlo technique using MCNPX computer program was adopted for the design work.In the first part of the work, the use of MCNPX for the shielding calculation was validated. A simple shielding experiment setup was performed using aluminum, stainless steel and lead as the shield. The experiment used Co-60 as the standard source and used NaI detector with gamma spectrometer for the counting system. The Co-60 attenuation coefficients of these materials were obtained from experiment and MCNPX calculation. Both results were compared and found to be in a very good agreement (the differences wereessentially less than 5%). The conclusion from this comparison was that the use of MCNPX for shielding calculation could provide quite accurate results. In the second part of the work, the preliminary design of new shield container was performed. Lead and stainless steel were both chosen as the materials for the container because they have good shielding properties. The shield container was to have two layers consisting of stainless steel and lead. Using both materials in the shield was foreseen to yield a good compromisebetween space utilization and cost. Moreover, it is preferable that stainless steel be the material in contact with the waste. This is because the waste is made of stainless steel and hence there should be no induced corrosion from material incompatibility. The design approachwas to vary the shield thickness in step in order to optimize the

  2. Experimental Shielding Evaluation of the Radiation Protection Provided by Residential Structures

    Science.gov (United States)

    Dickson, Elijah D.

    The human health and environmental effects following a postulated accidental release of radioactive material to the environment has been a public and regulatory concern since the early development of nuclear technology and researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to research and develop the technical basis for contemporary building shielding factors for the U.S. housing stock. Building shielding factors quantify the protection a certain building-type provides from ionizing radiation. 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 1950's era suburbia and is no longer applicable to the densely populated urban environments seen today. To analyze a building's radiation shielding properties, the ideal approach would be to subject a variety of building-types to various radioactive materials and measure the radiation levels in and around the building. While this is not entirely practicable, this research uniquely analyzes the shielding effectiveness of a variety of likely U.S. residential buildings from a realistic source term in a laboratory setting. Results produced in the investigation provide a comparison between theory and experiment behind building shielding factor methodology by applying laboratory measurements to detailed computational models. These models are used to develop a series of validated building shielding factors for generic residential housing units using the computational code MCNP5. For these building shielding factors to be useful in radiologic consequence assessments and emergency response planning, two types of shielding factors have been developed for; (1) the shielding effectiveness of each structure within a semi-infinite cloud of radioactive material, and (2) the shielding effectiveness of each structure

  3. Design of the precast, post-tensioned concrete shielding structure for the TFTR neutral beam test cell

    International Nuclear Information System (INIS)

    At the TFTR facility, the Neutral Beam Test Cell is a room separated from the TFTR Cell by a 4-foot-thick concrete wall and devoted to testing the neutral beam injector. The function of the shielding structure is to protect personnel from radiation casued by pulsing the injector. The distance from the TFTR device to the injector is large enough to permit use of magnetic materials in the shielding structure, and the neutron flux levels are small enough so that ordinary concrete of moderate thickness may be employed. Radiation considerations are not discussed in this paper, which is devoted to a description of the structural design of the shield

  4. Optimization of multi-layered metallic shield

    International Nuclear Information System (INIS)

    Research highlights: → We investigated the problem of optimization of a multi-layered metallic shield. → The maximum ballistic limit velocity is a criterion of optimization. → The sequence of materials and the thicknesses of layers in the shield are varied. → The general problem is reduced to the problem of Geometric Programming. → Analytical solutions are obtained for two- and three-layered shields. - Abstract: We investigate the problem of optimization of multi-layered metallic shield whereby the goal is to determine the sequence of materials and the thicknesses of the layers that provide the maximum ballistic limit velocity of the shield. Optimization is performed under the following constraints: fixed areal density of the shield, the upper bound on the total thickness of the shield and the bounds on the thicknesses of the plates manufactured from every material. The problem is reduced to the problem of Geometric Programming which can be solved numerically using known methods. For the most interesting in practice cases of two-layered and three-layered shields the solution is obtained in the explicit analytical form.

  5. ITER cryostat thermal shield detailed design

    International Nuclear Information System (INIS)

    The structural design and study on fabrication and assembly of the cryostat thermal shield for International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat thermal shield is attached to cover the cryostat inner wall in order to reduce the radiation heat loads applied to the superconducting coils operation at 4 K. The thermal shield consists of low-emissivity foils which are passively cooled and shield plates which are actively cooled with low temperature helium gas. The foils are multi-layered assemblies and are attached on both surfaces of the shield plates. The material of the foils are silver coated 304 stainless steel, polyimide or polyester. The silver coated stainless steel foils should be adopted to the foils at the locations where radiation dose is over 10 MGy. The route of coolant pipes for the shield plates is designed so as to keep the surface temperature of the shield plates below 100 K. This report describes the detailed design of the cryostat thermal shield, and outlines the fabrication and assembly procedures. (J.P.N.)

  6. ITER cryostat thermal shield detailed design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Hamada, Kazuya; Takahashi, Hiroyuki; Tada, Eisuke; Kato, Takashi [Department of Fusion Engineering Research, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Nishikawa, Akira

    1999-03-01

    The structural design and study on fabrication and assembly of the cryostat thermal shield for International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat thermal shield is attached to cover the cryostat inner wall in order to reduce the radiation heat loads applied to the superconducting coils operation at 4 K. The thermal shield consists of low-emissivity foils which are passively cooled and shield plates which are actively cooled with low temperature helium gas. The foils are multi-layered assemblies and are attached on both surfaces of the shield plates. The material of the foils are silver coated 304 stainless steel, polyimide or polyester. The silver coated stainless steel foils should be adopted to the foils at the locations where radiation dose is over 10 MGy. The route of coolant pipes for the shield plates is designed so as to keep the surface temperature of the shield plates below 100 K. This report describes the detailed design of the cryostat thermal shield, and outlines the fabrication and assembly procedures. (J.P.N.)

  7. Flexible shielding system for radiation protection

    Science.gov (United States)

    Babin, A.

    1972-01-01

    Modular construction of low cost flexible radiation shielding panels consists of water filled steels cans, zinc bromide windows, turntable unit, master-slave manipulators, and interlocking lead bricks. Easy modifications of shielding wall thicknesses are obtained by rearranging overall geometry of portable components.

  8. Study of the concrete shielding properties

    International Nuclear Information System (INIS)

    An analysis was performed of chemical composition, production technology and operating temperature influencing the shielding properties of a number of ordinary concrete types. Computation results of radiation transmission proved the significance of detailed knowledge of all these factors in the reactor shield production. (author)

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

    International Nuclear Information System (INIS)

    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)

  10. Radiation Shielding Systems Using Nanotechnology

    Science.gov (United States)

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

    2011-01-01

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

  11. Neutron shielding heat insulation material

    International Nuclear Information System (INIS)

    Purpose: To improve decceleration and absorption of neutrons by incorporating neutron moderators and neutron absorbers in asbestos to thereby increase hydrogen concentration. Constitution: A mixture consisting of crysotile asbestos, surface active agent and water is well stirred and compounded to open the crysotile asbestos filaments and prepare a high viscosity slurry. After adding hydroxides such as magnesium hydroxide, hydrated salts such as magnesium borate hydrate or water containing minerals such as alumina cement hydrate, or boron compound to the slurry, the slurry is charged in a predetermined die, and dried and compressed to prepare shielding heat insulation products. The crysotile asbestos has 18 - 15 wt.% of water of crystallinity in the structure and contains a considerably high hydrogen concentration that acts as neutron moderators. (Kawakami, Y.)

  12. Electromagnetic interference shielding effectiveness of monolayer graphene.

    Science.gov (United States)

    Hong, Seul Ki; Kim, Ki Yeong; Kim, Taek Yong; Kim, Jong Hoon; Park, Seong Wook; Kim, Joung Ho; Cho, Byung Jin

    2012-11-16

    We report the first experimental results on the electromagnetic interference (EMI) shielding effectiveness (SE) of monolayer graphene. The monolayer CVD graphene has an average SE value of 2.27 dB, corresponding to ~40% shielding of incident waves. CVD graphene shows more than seven times (in terms of dB) greater SE than gold film. The dominant mechanism is absorption rather than reflection, and the portion of absorption decreases with an increase in the number of graphene layers. Our modeling work shows that plane-wave theory for metal shielding is also applicable to graphene. The model predicts that ideal monolayer graphene can shield as much as 97.8% of EMI. This suggests the feasibility of manufacturing an ultrathin, transparent, and flexible EMI shield by single or few-layer graphene. PMID:23085718

  13. Radiation Shielding for Nuclear Thermal Propulsion

    Science.gov (United States)

    Caffrey, Jarvis A.

    2016-01-01

    Design and analysis of radiation shielding for nuclear thermal propulsion has continued at Marshall Space Flight Center. A set of optimization tools are in development, and strategies for shielding optimization will be discussed. Considerations for the concurrent design of internal and external shielding are likely required for a mass optimal shield design. The task of reducing radiation dose to crew from a nuclear engine is considered to be less challenging than the task of thermal mitigation for cryogenic propellant, especially considering the likely implementation of additional crew shielding for protection from solar particles and cosmic rays. Further consideration is thus made for the thermal effects of radiation absorption in cryogenic propellant. Materials challenges and possible methods of manufacturing are also discussed.

  14. Shield verification and validation action matrix summary

    Energy Technology Data Exchange (ETDEWEB)

    Boman, C.

    1992-02-01

    WSRC-RP-90-26, Certification Plan for Reactor Analysis Computer Codes, describes a series of action items to be completed for certification of reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. Validation and verification are integral part of the certification process. This document identifies the work performed and documentation generated to satisfy these action items for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system, it is not certification of the complete SHIELD system. Complete certification will follow at a later date. Each action item is discussed with the justification for its completion. Specific details of the work performed are not included in this document but can be found in the references. The validation and verification effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system computer code is completed.

  15. The use of nipple shields: A review

    Directory of Open Access Journals (Sweden)

    Selina eChow

    2015-10-01

    Full Text Available A nipple shield is a breastfeeding aid with a nipple-shaped shield that is positioned over the nipple and areola prior to nursing. Nipple shields are usually recommended to mothers with flat nipples or in cases in which there is a failure of the baby to effectively latch onto the breast within the first two days postpartum. The use of nipple shields is a controversial topic in the field of lactation. Its use has been an issue in the clinical literature since some older studies discovered reduced breast milk transfer when using nipple shields, while more recent studies reported successful breastfeeding outcomes. The purpose of this review was to examine the evidence and outcomes with nipple shield use. Methods: A literature search was conducted in Ovid MEDLINE, OLDMEDLINE, EMBASE Classic, EMBASE, Cochrane Central Register of Controlled Trials and CINAHL. The primary endpoint was any breastfeeding outcome following nipple shield use. Secondary endpoints included the reasons for nipple shield use and the average/median length of use. For the analysis, we examined the effect of nipple shield use on physiological responses, premature infants, mothers’ experiences, and health professionals’ experiences. Results: The literature search yielded 261 articles, 14 of which were included in this review. Of these 14 articles, three reported on physiological responses, two reported on premature infants, eight reported on mothers’ experiences, and one reported on health professionals’ experiences. Conclusion: Through examining the use of nipple shields, further insight is provided on the advantages and disadvantages of this practice, thus allowing clinicians and researchers to address improvements on areas that will benefit mothers and infants the most.

  16. DEMONSTRATION BULK VITRIFICATION SYSTEM (DBVS) EXTERNAL REVIEW

    International Nuclear Information System (INIS)

    The Hanford mission to retrieve and immobilize 53 million gallons of radioactive waste from 177 underground storage tanks will be accomplished using a combination of processing by the waste treatment plant currently under construction, and a supplemental treatment that would process low-activity waste. Under consideration for this treatment is bulk vitrification, a versatile joule-heated melter technology which could be deployed in the tank farms. The Department proposes to demonstrate this technology under a Research, Development and Demonstration (RD and D) permit issued by the Washington State Department of Ecology using both non-radioactive simulant and blends of actual tank waste. From the demonstration program, data would be obtained on cost and technical performance to enable a decision on the potential use of bulk vitrification as the supplemental treatment technology for Hanford. An independent review by sixteen subject matter experts was conducted to assure that the technical basis of the demonstration facility design would be adequate to meet the objectives of the Demonstration Bulk Vitrification System (DBVS) program. This review explored all aspects of the program, including flowsheet chemistry, project risk, vitrification, equipment design and nuclear safety, and was carried out at a time when issues can be identified and corrected. This paper describes the mission need, review approach, technical recommendations and follow-on activities for the DBVS program

  17. Important Cautions in Shielding Computation by using the FLUKA Code

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Seung Uk; Oh, Sunju; Song, Yong keun; Kum, Oyeon [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Nam, Sang Hee [Inje Univ.., Kimhae (Korea, Republic of)

    2014-10-15

    It is important to consider how efficiently the facility is designed against the shielding and activation problems in order to minimize the radiation exposure. In order to evaluate shielding capability of facility, although the simple calculation with approximate methods were popular until recently, owe to the development of information technology and the advances of computational mathematics, the Monte Carlo codes such as the MCNP, FLUKA, GEANT, and PHITS which can provide accurate answer are popularly used now. Advantage of Monte Carlo code is to perform the correct calculation but takes a long time for computing. More importantly, the exact and precise input data for Monte Carlo codes is essential in order to obtain accurate results. Thus, in this paper, important cautions are presented for shielding computation with the FLUKA code because the ignorance of such important cautions makes big troubles. The absorbed doses and errors show similar tendency in the comparison of groups, 1 and 2. Specifically, the results confirmed the more similar tendency in the high convergence areas. In group 3, although the comparison with groups, 1 and 2, shows the similar absorbed dose in the detectors with high convergences, the results themselves are unreliable because the errors are 99.9%. Thus, we need more careful attention to the average value and error in using the FLUKA code. Simply, it is better for us to have other benchmark tools such as MCNPX. However, it is recommended that the best computing method with the FLUKA code is the same as the computing of group 2, the usual multiprocessing with semi-automatic data handling. As shown in group 3, higher number of the cycle is a better method than the higher history to get more reliable result or to reduce errors. However, these values should be carefully evaluated.

  18. Improved Metal-Polymeric Laminate Radiation Shielding Project

    Data.gov (United States)

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

  19. Foam-Reinforced Polymer Matrix Composite Radiation Shields Project

    Data.gov (United States)

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

  20. A User's Manual for the NRN Shield Design Method

    International Nuclear Information System (INIS)

    This report describes a code system for bulk shield design written for a Ferranti Mercury computer and is intended as a manual for those using the programme. The idea of an 'almost direct' flux, as in the removal theory serves as a basis for further development of the theory. An important aspiration has been to minimize the manual work of administering the codes. The codes concerned are: NECO, computing necessary group constants from primary data, REFUSE and REBOX (infinite plane or cylindrical, and box geometry, respectively), computing removal flux, NEDI a one-dimensional (plane, spherical, cylindrical) diffusion multigroup code, and SALOME a Monte Carlo code computing the gamma flux. Output tapes are constructed for direct use as input tapes, when required, for a following code

  1. MCNP benchmark calculation: GCFR grid-plate shield design, configuration II.A

    International Nuclear Information System (INIS)

    This report describes the Monte Carlo MCNP analysis of one of the GCFR Shield Design experimental configurations which has been constructed and analyzed at the Test Shielding Facility in ORNL. It is a part of the benchmarking program for MCNP, which has been agreed upon with HRB, Mannheim. The calculated response results for the selected detectors agree within 10 % with the measured ones, what can be considered as a very good agreement. The code appears to be a reliable tool for the analysis of similar systems. (author)

  2. Radiation shielding calculations for MuCool Test Area at Fermilab

    CERN Document Server

    Rakhno, I

    2004-01-01

    The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this shielding study the results of Monte Carlo radiation shielding calculations performed using the MARS14 code for the MuCool Test Area and including the downstream portion of the target hall and berm around it, access pit, service building, and parking lot are presented and discussed within the context of the proposed MTA experimental configuration.

  3. Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

    CERN Document Server

    Newhauser, W D; 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.

  4. Radiation shielding calculations for MuCool test area at Fermilab

    Energy Technology Data Exchange (ETDEWEB)

    Igor Rakhno; Carol Johnstone

    2004-05-26

    The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this shielding study the results of Monte Carlo radiation shielding calculations performed using the MARS14 code for the MuCool Test Area and including the downstream portion of the target hall and berm around it, access pit, service building, and parking lot are presented and discussed within the context of the proposed MTA experimental configuration.

  5. AAPM Task Group 108: PET and PET/CT Shielding Requirements

    International Nuclear Information System (INIS)

    The shielding of positron emission tomography (PET) and PET/CT (computed tomography) facilities presents special challenges. The 0.511 MeV annihilation photons associated with positron decay are much higher energy than other diagnostic radiations. As a result, barrier shielding may be required in floors and ceilings as well as adjacent walls. Since the patient becomes the radioactive source after the radiopharmaceutical has been administered, one has to consider the entire time that the subject remains in the clinic. In this report we present methods for estimating the shielding requirements for PET and PET/CT facilities. Information about the physical properties of the most commonly used clinical PET radionuclides is summarized, although the report primarily refers to fluorine-18. Typical PET imaging protocols are reviewed and exposure rates from patients are estimated including self-attenuation by body tissues and physical decay of the radionuclide. Examples of barrier calculations are presented for controlled and noncontrolled areas. Shielding for adjacent rooms with scintillation cameras is also discussed. Tables and graphs of estimated transmission factors for lead, steel, and concrete at 0.511 MeV are also included. Meeting the regulatory limits for uncontrolled areas can be an expensive proposition. Careful planning with the equipment vendor, facility architect, and a qualified medical physicist is necessary to produce a cost effective design while maintaining radiation safety standards

  6. AAPM Task Group 108: PET and PET/CT shielding requirements.

    Science.gov (United States)

    Madsen, Mark T; Anderson, Jon A; Halama, James R; Kleck, Jeff; Simpkin, Douglas J; Votaw, John R; Wendt, Richard E; Williams, Lawrence E; Yester, Michael V

    2006-01-01

    The shielding of positron emission tomography (PET) and PET/CT (computed tomography) facilities presents special challenges. The 0.511 MeV annihilation photons associated with positron decay are much higher energy than other diagnostic radiations. As a result, barrier shielding may be required in floors and ceilings as well as adjacent walls. Since the patient becomes the radioactive source after the radiopharmaceutical has been administered, one has to consider the entire time that the subject remains in the clinic. In this report we present methods for estimating the shielding requirements for PET and PET/CT facilities. Information about the physical properties of the most commonly used clinical PET radionuclides is summarized, although the report primarily refers to fluorine-18. Typical PET imaging protocols are reviewed and exposure rates from patients are estimated including self-attenuation by body tissues and physical decay of the radionuclide. Examples of barrier calculations are presented for controlled and noncontrolled areas. Shielding for adjacent rooms with scintillation cameras is also discussed. Tables and graphs of estimated transmission factors for lead, steel, and concrete at 0.511 MeV are also included. Meeting the regulatory limits for uncontrolled areas can be an expensive proposition. Careful planning with the equipment vendor, facility architect, and a qualified medical physicist is necessary to produce a cost effective design while maintaining radiation safety standards. PMID:16485403

  7. Qualification of the FENDL neutron cross-sections based on bulk shielding experiments

    Energy Technology Data Exchange (ETDEWEB)

    Santamarina, A.; Benmansour, L.; Gastaldi, B.; Jacqmin, R.; Camous, B.; Philibert, H. [C.E.A. Cadarache, Saint-Paul-lez-Durance (France). DRN/DER/SPRC

    1997-08-01

    The main objective of this work is to produce an improved evaluation of the Fe56 nuclear data tailored to the requirements of the engineering and development activities (EDA) of the international thermonuclear experimental reactor (ITER). This improved data, obtained through sensitivity studies and a `trend analysis` method, is intended for the final FENDL-2 library. (orig.) 13 refs.

  8. Qualification of the FENDL neutron cross-sections based on bulk shielding experiments

    International Nuclear Information System (INIS)

    The main objective of this work is to produce an improved evaluation of the Fe56 nuclear data tailored to the requirements of the engineering and development activities (EDA) of the international thermonuclear experimental reactor (ITER). This improved data, obtained through sensitivity studies and a 'trend analysis' method, is intended for the final FENDL-2 library. (orig.)

  9. Benchmarking the multipole shielding polarizability/reaction field approach to solvation against QM/MM: Applications to the shielding constants of N-methylacetamide

    Science.gov (United States)

    Kjær, Hanna; Sauer, Stephan P. A.; Kongsted, Jacob

    2011-01-01

    We present a benchmark study of a combined multipole shielding polarizability/reaction field (MSP/RF) approach to the calculation of both specific and bulk solvation effects on nuclear magnetic shielding constants of solvated molecules. The MSP/RF scheme is defined by an expansion of the shielding constants of the solvated molecule in terms of electric field and field gradient property derivatives derived from single molecule ab initio calculations. The solvent electric field and electric field gradient are calculated based on data derived from molecular dynamics simulations, thereby accounting for solute-solvent dynamical effects. The MSP/RF method is benchmarked against polarizable quantum mechanics/molecular mechanics (QM/MM) calculations. The best agreement between the MSP/RF and QM/MM approaches is found by truncating the electric field expansion in the MSP/RF approach at the linear electric field level which is due to the cancelation of errors. In addition, we investigate the sensitivity of the results due to the choice of one-electron basis set in the ab initio calculations of the property derivatives and find that these derivatives are affected by the basis set in a way similar to the shielding constants themselves.

  10. Extraterrestrial Regolith Derived Atmospheric Entry Heat Shields

    Science.gov (United States)

    Hogue, Michael D.; Mueller, Robert P.; Sibille, Laurent; Hintze, Paul E.; Rasky, Daniel J.

    2016-01-01

    High-mass planetary surface access is one of NASAs technical challenges involving entry, descent and landing (EDL). During the entry and descent phase, frictional interaction with the planetary atmosphere causes a heat build-up to occur on the spacecraft, which will rapidly destroy it if a heat shield is not used. However, the heat shield incurs a mass penalty because it must be launched from Earth with the spacecraft, thus consuming a lot of precious propellant. This NASA Innovative Advanced Concept (NIAC) project investigated an approach to provide heat shield protection to spacecraft after launch and prior to each EDL thus potentially realizing significant launch mass savings. Heat shields fabricated in situ can provide a thermal-protection system for spacecraft that routinely enter a planetary atmosphere. By fabricating the heat shield with space resources from materials available on moons and asteroids, it is possible to avoid launching the heat-shield mass from Earth. Regolith has extremely good insulating properties and the silicates it contains can be used in the fabrication and molding of thermal-protection materials. In this paper, we will describe three types of in situ fabrication methods for heat shields and the testing performed to determine feasibility of this approach.

  11. Shielding pebble transfer system for thermonuclear device

    International Nuclear Information System (INIS)

    In a system for supplying shielding pebbles to a vacuum vessel filled with the shielding pebbles in a gap of a double-walled structure, a supply port for the shielding pebbles is formed in a diverging shape, and a corny object is disposed at the center of the flow channel, or protrusions are formed in the vicinity of the supply port. Alternatively, a small object is disposed at the center of the flow channel of the supply port, and the small object is supported swingably and tiltably by elastic members. In addition, the upper plate of the vacuum vessel is slanted having the supply port of the shielding pebbles as a top, and a slanting angle relative to a horizontal axis is made greater than the resting angle of the shielding pebble accumulation layer. The shielding pebbles are jetted out from the supply port and spread to the peripheries, abut against the inner surface of the vacuum vessel, jump up and then accumulate. Accordingly, they can be accumulated dispersingly without being localized. An uniform accumulation layer is obtained to form a vacuum vessel having uniform and high shielding performance. (N.H.)

  12. Shielding options for the ITER conceptual design

    International Nuclear Information System (INIS)

    Several shield options were analyzed for the ITER conceptual design to minimize the nuclear responses in the toroidal field (TF) coils. The total nuclear heating in the physics phase and the insulator dose in the technology phase are the most critical parameters in the design process. The first shield option has type 316 stainless steel and water shielding material. Steel and water also serve as structural material and coolant, respectively. The second option is similar to the first except that borated water is used instead of ordinary water. The other two options include a small layer of lead or boron carbide (B4C) at the back of the shield. The last three shield options were considered to reduce the nuclear heating in the toroidal field coils relative to the steel/water shield. An optimization process was performed taking into consideration the thermal-hydraulics and the engineering- requirements to define the shield configuration. A careful integration was performed to calculate the total nuclear heating in the toroidal field coils which account for the neutron wall loading distribution, the change in the shield thickness in the poloidal direction, and the space between the toroidal field coils in the divertor zone. The results show that the steel/water/Pb and the steel/borated water shield options are very close in terms of the total nuclear heating in the toroidal field coils and the dose in the insulator material. The other two options, steel/water and steel/water/B4C deposit more nuclear heating in the toroidal field coils. 5 refs., 3 figs., 5 tabs

  13. Radiation protection, radiation safety and radiation shielding assessment of HIE-ISOLDE

    International Nuclear Information System (INIS)

    The high intensity and energy ISOLDE (HIE-ISOLDE) project is an upgrade to the existing ISOLDE facility at CERN. The foreseen increase in the nominal intensity and the energy of the primary proton beam of the existing ISOLDE facility aims at increasing the intensity of the produced radioactive ion beams (RIBs). The currently existing ISOLDE facility uses the proton beam from the proton-synchrotron booster with an energy of 1.4 GeV and an intensity up to 2 μA. After upgrade (final stage), the HIE-ISOLDE facility is supposed to run at an energy up to 2 GeV and an intensity up to 4 μA. The foreseen upgrade imposes constrains, from the radiation protection and the radiation safety point of view, to the existing experimental and supply areas. Taking into account the upgraded energy and intensity of the primary proton beam, a new assessment of the radiation protection and radiation safety of the HIE-ISOLDE facility is necessary. Special attention must be devoted to the shielding assessment of the beam dumps and of the experimental areas. In this work the state-of-the-art Monte Carlo particle transport simulation program FLUKA was used to perform the computation of the ambient dose equivalent rate distribution and of the particle fluxes in the projected HIE-ISOLDE facility (taking into account the upgrade nominal primary proton beam energy and intensity) and the shielding assessment of the facility, with the aim of identifying in the existing facility (ISOLDE) the critical areas and locations where new or reinforced shielding may be necessary. The consequences of the upgraded proton beam parameters on the operational radiation protection of the facility were studied. (authors)

  14. The use of radiological characterisation in support of the design and build of a new facility in an area of elevated dose rate - 16009

    International Nuclear Information System (INIS)

    A new building, a maintenance facility, is to be constructed in the 'separation area' of the Sellafield Site. Sellafield is a complex and busy nuclear facility covering about two square miles in the north-west of the United Kingdom. The facility, which is to provide necessary storage and maintenance functionality to support bulk waste retrievals from legacy silos, is being built in close proximity to spent fuel storage ponds, intermediate level legacy waste stores and a pipe-bridge carrying active materials. In addition, the site is adjacent to a main pedestrian and vehicle thoroughfare and a railway line used for the regular transfer of nuclear materials. The facility itself is to be built on the site of a recently demolished active facility. The ambient gamma dose rates in the construction area have been measured to be typically 5 to 50 μSv/hr. Although there are known specific sources of dose rate close to the construction site, the relative contributions of each within the envelope of the new facility are unknown. This paper describes how a series of gamma dose rate measurements, gamma spectroscopy measurements and gamma-ray imaging surveys, using RadScanR 800, have been used to better understand the origins of the dose rates at a number of key locations within the area. The results of the survey are being used to assess the requirements for shielding within the structure of the facility in order to reduce the dose to personnel that will work there when it is operational. The results are also being used to identify whether any localised shielding would prove beneficial for reduction of the dose both during construction and occupancy. The unique mix of facilities surrounding the site have meant that the contributions to the dose rate come not only from adjacent facilities in the form of unscattered, higher energy, penetrating radiation, but also in the form of lower energy scattered radiation, both from radiation that has passed through shielding and also in

  15. Carbon nanostructure composite for electromagnetic interference shielding

    Indian Academy of Sciences (India)

    Anupama Joshi; Suwarna Datar

    2015-06-01

    This communication reviews current developments in carbon nanostructure-based composite materials for electromagnetic interference (EMI) shielding. With more and more electronic gadgets being used at different frequencies, there is a need for shielding them from one another to avoid interference. Conventionally, metal-based shielding materials have been used. But due to the requirement of light weight, corrosion resistive materials, lot of work is being done on composite materials. In this research the forerunner is the nanocarbon-based composite material whose different forms add different characteristics to the composite. The article focusses on composites based on graphene, graphene oxide, carbon nanotubes, and several other novel forms of carbon.

  16. Safety assessment of ETRR-2 shielding

    International Nuclear Information System (INIS)

    The ETRR-2 is the second Egyptian research reactor. The reactor is of an open pool type and shielding and dose calculation is one of the most important scopes of the reactor safety. The core radiation source has been calculated using the Madland-Nix model (MNM) for a prompt fission neutron spectrum. The ANISN code has been used to determine the flux and dose through the axial and radial layers of the reactor shielding. It has been found that the calculated dose in the outer area of the reactor shield does not exceed the maximum allowable dose level which is in a good agreement with the measurement. (author)

  17. Gamma Ray Shielding from Saudi White Sand

    OpenAIRE

    Al-horayess OKLA; Al-Dayel OMAR; Hefne JAMEEL; Al-Ajyan TURKI; Bagazi ALI

    2010-01-01

    This study is a comparison of gamma ray linear attenuation coefficient of two typs of shielding materials made of Saudi white and red sand. Each shield was consisted of one part of cement two parts of sand in addi-tion to water. Different thicknesses were tested. The concentrations of all elements in each shield material were determined by Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The results obtained from the ICP-MS were used in MCNP4B (Monte Carlo N-Particle Transport Computer ...

  18. Planetary surface reactor shielding using indigenous materials

    International Nuclear Information System (INIS)

    The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials

  19. Mining the bulk positron lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Aourag, H.; Guittom, A. [Centre de Recherche Nucleaire d' Alger (CRNA), Alger Gare - Algiers (Algeria)

    2009-02-15

    We introduce a new approach to investigate the bulk positron lifetimes of new systems based on data-mining techniques. Through data mining of bulk positron lifetimes, we demonstrate the ability to predict the positron lifetimes of new semiconductors on the basis of available semiconductor data already studied. Informatics techniques have been applied to bulk positron lifetimes for different tetrahedrally bounded semiconductors in order to discover computational design rules. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  20. Mining the bulk positron lifetime

    International Nuclear Information System (INIS)

    We introduce a new approach to investigate the bulk positron lifetimes of new systems based on data-mining techniques. Through data mining of bulk positron lifetimes, we demonstrate the ability to predict the positron lifetimes of new semiconductors on the basis of available semiconductor data already studied. Informatics techniques have been applied to bulk positron lifetimes for different tetrahedrally bounded semiconductors in order to discover computational design rules. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. Advances in bulk port development

    Energy Technology Data Exchange (ETDEWEB)

    Soros, P. (Soros Associates Consulting Engineers, New York, NY (USA))

    1991-03-01

    The article features several recently developed bulk ports which illustrate aspects of new technology or concepts in maritime transport. Low handling capacity bulk terminals at Ponta da Madeira, Brazil and Kooragang Island, Australia and the low-cost bulk port at Port of Corpus Christi, Texas are described. Operations at the ports of Pecket and Tocopilla in Chile, which had special technical problems, are mentioned. Coal terminals at Port Kembla, Australia and St. Johns River in Florid Jacksonville, Florida are featured as examples of terminals which had to be designed to meet high environmental standards. 13 refs., 2 figs., 14 photos.

  2. Thermal design of top shield

    International Nuclear Information System (INIS)

    Full text of publication follows: Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type fast reactor. The top shield forms the top cover for the main vessel (MV) and includes roof slab (RS), large rotatable plug (LRP), small rotatable plug (SRP) and control Plug (CP). RS, LRP and SRP are box type structures consisting of top and bottom plates stiffened by radial stiffeners and vertical penetration shells. TS is exposed to argon cover gas provided above sodium pool on the bottom side and reactor containment building air at the top. Heat transfer takes place through the argon cover gas to the bottom plate of TS. Annular gaps are formed between the components supported on TS and the component penetrations through which cellular convection takes place. A single thermal shield provided below TS reduces the heat flux to the bottom plate to 1.15 kW/m2. The MV (SS 316 LN) is welded to RS (carbon steel A48 P2) through a dissimilar metal weld. A step in RS and an anti convection barrier (ACB) outside RS are provided to limit the temperature at the MV-RS junction. The MV is surrounded by safety vessel (SV) and reactor vault made of concrete. Thermal insulation is provided outside SV to limit the heat transfer to the reactor vault. The design requirements of TS are to maintain the operating temperature at 383-393 K, limit the temperature difference (ΔT) across the height of TS to 20 / 100 K under normal operation/loss of cooling, provide minimum annular gap size at the component penetrations, provide a nearly linear temperature gradient in the CP portion within the height of TS, maintain the temperature of top plate of CP > 383 K, limit the ΔT across the top plate of CP to 2 K, limit the temperature near the inflatable / backup seal to 393 K, limit the temperature at the MV-RS junction and the heat flux to the reactor vault. The total heat transferred to TS is estimated to be 210 kW. A dedicated closed loop cooling system with a total flow rate of 10 m

  3. Brachytherapy structural shielding calculations using Monte Carlo generated, monoenergetic data

    International Nuclear Information System (INIS)

    Purpose: To provide a method for calculating the transmission of any broad photon beam with a known energy spectrum in the range of 20–1090 keV, through concrete and lead, based on the superposition of corresponding monoenergetic data obtained from Monte Carlo simulation. Methods: MCNP5 was used to calculate broad photon beam transmission data through varying thickness of lead and concrete, for monoenergetic point sources of energy in the range pertinent to brachytherapy (20–1090 keV, in 10 keV intervals). The three parameter empirical model introduced byArcher et al. [“Diagnostic x-ray shielding design based on an empirical model of photon attenuation,” Health Phys. 44, 507–517 (1983)] was used to describe the transmission curve for each of the 216 energy-material combinations. These three parameters, and hence the transmission curve, for any polyenergetic spectrum can then be obtained by superposition along the lines of Kharrati et al. [“Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities,” Med. Phys. 34, 1398–1404 (2007)]. A simple program, incorporating a graphical user interface, was developed to facilitate the superposition of monoenergetic data, the graphical and tabular display of broad photon beam transmission curves, and the calculation of material thickness required for a given transmission from these curves. Results: Polyenergetic broad photon beam transmission curves of this work, calculated from the superposition of monoenergetic data, are compared to corresponding results in the literature. A good agreement is observed with results in the literature obtained from Monte Carlo simulations for the photon spectra emitted from bare point sources of various radionuclides. Differences are observed with corresponding results in the literature for x-ray spectra at various tube potentials, mainly due to the different broad beam conditions or x-ray spectra assumed. Conclusions

  4. Bulk Nuclear Properties from Reactions

    OpenAIRE

    Danielewicz, P.

    2002-01-01

    Extraction of bulk nuclear properties by comparing reaction observables to results from semiclassical transport-model simulations is discussed. Specific properties include the nuclear viscosity, incompressibility and constraints on the nuclear pressure at supranormal densities.

  5. Bulk charges in eleven dimensions

    CERN Document Server

    Hawking, Stephen William

    1998-01-01

    Eleven dimensional supergravity has electric type currents arising from the Chern-Simon and anomaly terms in the action. However the bulk charge integrates to zero for asymptotically flat solutions with topological trivial spatial sections. We show that by relaxing the boundary conditions to generalisations of the ALE and ALF boundary conditions in four dimensions one can obtain static solutions with a bulk charge preserving between 1/16 and 1/4 of the supersymmetries. One can introduce membranes with the same sign of charge into these backgrounds. This raises the possibility that these generalized membranes might decay quantum mechanically to leave just a bulk distribution of charge. Alternatively and more probably, a bulk distribution of charge can decay into a collection of singlely charged membranes. Dimensional reductions of these solutions lead to novel representations of extreme black holes in four dimensions with up to four charges. We discuss how the eleven-dimensional Kaluza-Klein monopole wrapped a...

  6. Shielded ADR Magnets For Space Applications Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Phase II program will concentrate on manufacturing of qualified low-current, light-weight, 10K ADR magnets for space application. Shielded ADR solenoidal...

  7. Shielding benchmark test for JENDL-3T

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira (Japan Atomic Energy Research Inst., Tokai, Ibaraki. Tokai Research Establishment)

    1988-03-01

    The results of the shielding benchmark tests for JENDL-3T (testing stage version of JENDL-3), performed by JNDC Shielding Sub-working group, are summarized. Especially, problems of total cross-section in MeV range for O, Na, Fe, revealed from the analysis of the Broomstick's experiment, are discussed in details. For the deep penetration profiles of Fe, which is very important feature in shielding calculation, ASPIS benchmark experiment is analysed and discussed. From the study overall applicability of JENDL-3T data for the shielding calculation is confirmed. At the same time some problems still remained are also pointed out. By the reflection of this feedback information applicability of JENDL-3, forth coming official version, will be greatly improved.

  8. Shielded ADR Magnets For Space Applications Project

    Data.gov (United States)

    National Aeronautics and Space Administration — An important consideration of the use of superconducting magnets in ADR applications is shielding of the other instruments in the vicinity of the superconducting...

  9. Long Duration Space Shelter Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Physical Sciences Inc. (PSI) has developed a ceramic composite material system that is more effective for shielding both GCR and SPE than aluminum. The composite...

  10. Long Duration Space Shelter Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Physical Sciences Inc. (PSI) has developed fiber reinforced ceramic composites for radiation shielding that can be used for external walls in long duration manned...

  11. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  12. Specification for lead bricks for radiation shielding

    International Nuclear Information System (INIS)

    Specification with metric dimensions for two systems of interlocking lead bricks for building permanent or temporary shielding walls including numbering system with illustrations, and schedule for ordering purposes. (author)

  13. Passive Magnetic Shielding in Gradient Fields

    CERN Document Server

    Bidinosti, C P

    2013-01-01

    The effect of passive magnetic shielding on dc magnetic field gradients imposed by both external and internal sources is studied. It is found that for concentric cylindrical or spherical shells of high permeability material, higher order multipoles in the magnetic field are shielded progressively better, by a factor related to the order of the multipole. In regard to the design of internal coil systems for the generation of uniform internal fields, we show how one can take advantage of the coupling of the coils to the innermost magnetic shield to further optimize the uniformity of the field. These results demonstrate quantitatively a phenomenon that was previously well-known qualitatively: that the resultant magnetic field within a passively magnetically shielded region can be much more uniform than the applied magnetic field itself. Furthermore we provide formulae relevant to active magnetic compensation systems which attempt to stabilize the interior fields by sensing and cancelling the exterior fields clos...

  14. Limiting currents in shielded source configurations

    International Nuclear Information System (INIS)

    Limiting currents for laminar flow equilibria of relativistic electron beams in shielded source configurations are discussed. Results are presented for the constant applied magnetic field case, and for the case of constant beam radius

  15. Demonstration study on shielding safety analysis code. 7

    International Nuclear Information System (INIS)

    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 study 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) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a 3He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

  16. Nuclear waste packaging facility

    International Nuclear Information System (INIS)

    A nuclear waste packaging facility comprising: (a) a first section substantially surrounded by radiation shielding, including means for remotely handling waste delivered to the first section and for placing the waste into a disposal module; (b) a second section substantially surrounded by radiation shielding, including means for handling a deformable container bearing waste delivered to the second section, the handling means including a compactor and means for placing the waste bearing deformable container into the compactor, the compactor capable of applying a compacting force to the waste bearing containers sufficient to inelastically deform the waste and container, and means for delivering the deformed waste bearing containers to a disposal module; (c) a module transportation and loading section disposed between the first and second sections including a means for handling empty modules delivered to the facility and for loading the empty modules on the transport means; the transport means moving empty disposal modules to the first section and empty disposal modules to the second section for locating empty modules in a position for loading with nuclear waste, and (d) a grouting station comprising means for pouring grout into the waste bearing disposal module, and a capping station comprising means for placing a lid onto the waste bearing grout-filled disposal module to completely encapsulate the waste

  17. Electromagnetic shielding with polypyrrole-coated fabrics

    OpenAIRE

    Avloni, J.; L. De Florio; Henn, A. R.; R. Lau; Ouyang, M.; Sparavigna, A.

    2006-01-01

    Several shielding applications, to protect human health and electronic devices against dangerous effects of electromagnetic radiation, require solutions that fabrics can suitably fulfill. Here, we will investigate the electromagnetic interference shielding effectiveness of polypyrrole-coated polyester textiles, in the frequency range 100-1000 MHz. Insertion losses for several conductive fabrics with different surface resistivity ranging from 40 Ohm till the very low value of 3 Ohm were evalua...

  18. Evaluation of tube shielding; Utvaerdering av tubskyddsmaterial

    Energy Technology Data Exchange (ETDEWEB)

    Hjoernhede, Anders; Westberg, Stig-Bjoern; Henderson, Pamela; Wetterstroem, Jonas; Jonasson, Anna

    2007-12-15

    Problems with soot-blowing have increased recently because of the poor fuel quality. Studies show that removing all the deposit by soot-blowing increases the metal loss of the superheaters, which drastically shortens component lifetimes. A simple, effective and common way of increasing the lifetime is to use tube shielding. Austenitic stainless steels seem to be the type of material most commonly used for tube shielding. It is thought that they give better protection against material removal than ferritic steels, but the cost of austenitics is several times greater than ferritic steels. It is clear that there is a significant economic advantage in choosing the right material for tube shielding, even though it might be expected that the cheaper materials do not perform as well as the more expensive ones. The reason for the study reported here is that very little material data exists in the literature. Few, if any tests have been performed to study the choice of material for tube shielding. The goal was to compare and evaluate a number of materials in a boiler to see if it is possible to replace the shielding material presently used with cheaper alternatives. About a dozen different shielding materials were installed and exposed for 4000 hours on primary- and secondary superheaters in a waste-fired boiler in Norrkoeping (Haendeloe Boiler 14.75MW). In total, 130 m of test material were installed and measured in several positions: a least 150 thickness measurements, before and after, were made on every tube shield. The results showed that the greatest attack was found on the secondary superheater shielding, where both the gas- and steam temperatures were higher. When considering cost and lifetime Sicromal 10 and 12 (however not Sicromal 8) and 15Mo3 are recommended as being better than 253 MA. The results should be of interest to most plants firing biomass or waste

  19. Ablating and charring of heat shield materials

    Energy Technology Data Exchange (ETDEWEB)

    Rahimian, M.H.; Shabani, M.R. [Univ. of Tehran, Faculty of Engineering, Mechanical Engineering Dept., Tehran (Iran, Islamic Republic of)]. E-mail: rahimyan@ut.ac.ir; shubani@me.ut.ac.ir

    2003-07-01

    The objective of this research is to estimate ablating and charring of heat shield materials in severe aero thermal / erosive environments. This requires an accurate and rapid technique for its serious heat transfer with moving boundary. Aerodynamic heating is obtained by an explicit relation. Fully implicit method is used for heat transfer calculation. Moving boundary is captured by VOF method. Thickness of heat shield, temperature of moving surface and radiation heat is presented. The results are in good agreement with other calculations. (author)

  20. Ablating and charring of heat shield materials

    International Nuclear Information System (INIS)

    The objective of this research is to estimate ablating and charring of heat shield materials in severe aero thermal / erosive environments. This requires an accurate and rapid technique for its serious heat transfer with moving boundary. Aerodynamic heating is obtained by an explicit relation. Fully implicit method is used for heat transfer calculation. Moving boundary is captured by VOF method. Thickness of heat shield, temperature of moving surface and radiation heat is presented. The results are in good agreement with other calculations. (author)