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Sample records for detectors sampling shielding

  1. HPGe detector shielding adjustment

    International Nuclear Information System (INIS)

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  2. Detectors, sampling, shielding, and electronics for positron emission tomography

    International Nuclear Information System (INIS)

    Derenzo, S.E.

    1981-08-01

    A brief discussion of the important design elements for positron emission tomographs is presented. The conclusions are that the instrumentation can be improved by the use of larger numbers of small, efficient detectors closely packed in many rings, the development of new detector materials, and novel electronic designs to reduce the deadtime and increase maximum event rates

  3. Shielded regenerative neutron detector

    International Nuclear Information System (INIS)

    Terhune, J.H.; Neissel, J.P.

    1978-01-01

    An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

  4. Shielding and grounding in large detectors

    International Nuclear Information System (INIS)

    Radeka, V.

    1998-09-01

    Prevention of electromagnetic interference (EMI), or ''noise pickup,'' is an important design aspect in large detectors in accelerator environments. Shielding effectiveness as a function of shield thickness and conductivity vs the type and frequency of the interference field is described. Noise induced in transmission lines by ground loop driven currents in the shield is evaluated and the importance of low shield resistance is emphasized. Some measures for prevention of ground loops and isolation of detector-readout systems are discussed

  5. Electronically shielded solid state charged particle detector

    International Nuclear Information System (INIS)

    Balmer, D.K.; Haverty, T.W.; Nordin, C.W.; Tyree, W.H.

    1996-01-01

    An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite. 1 fig

  6. {sup 3}He detector analysis of some special shielding materials

    Energy Technology Data Exchange (ETDEWEB)

    Avdic, S; Pesic, M [Boris Kidric, Institute of Nuclear Sciences, Beograd (Yugoslavia); Marinkovic, P [ETF Belgrade Univ. (Yugoslavia)

    1990-07-01

    The shielding properties of commercial materials of reactor Experiments, Inc. (R/X) were analyzed at the facility which includes bare heavy water experimental reactor RB with external neutron converter ENC, The fast neutron spectrum measurements in energy range from 1 MeV to 10 MeV was performed using ORTEC semiconductor neutron detector with He{sup 3} in diode coincidence arrangement. The neutron spectra have been evaluated from measured pulse-height distribution using numerical code HE3 for computation of detector efficiency in a collimated neutron beam. The neutron dose rates behind ENC with and without sample R/X material were determined using cubic spline interpolation routine for calculating the corresponding flux-dose rate conversion factors. Satisfactory shielding properties of the examined material in a fast neutron field in measurements and calculations are demonstrated. (author)

  7. Shielded scanning electron microscope for radioactive samples

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. A detector for use in high energy bremsstrahlung shielding studies

    International Nuclear Information System (INIS)

    Wilson, O.J.; Thomson, J.E.M.

    1983-01-01

    The design, development and calibration of a detector based on the principle of the Moxon-Rae detector is discussed. It is ideally suited to the measurement of the energy fluence of photons transmitted through a thick shield which has been irradiated with high energy bremsstrahlung. The detection sensitivity is 10 4 to 10 5 times that of the P2 ion chamber

  9. Configuration Design of Detector Shielding for Gamma Prompt Analysis

    International Nuclear Information System (INIS)

    Elin-Nuraini; Darsono; Elisabeth

    2000-01-01

    Configuration on design of detector shielding for gamma prompt analysishas been performed. The aim of this design is to obtain effective shieldingmaterial and configuration that able to protect the detector for fastneutron. The result shown that detector shielding configuration that obtainedby configuration of water and concrete, would be able to absorb fast neutronup to 99.5 %. The neutron flux that passed through shielding configuration is2.4 x 10 3 n/cm 2 dt, in the detector position of 60 cm (forward neutron beamdirection) on the X axis and 30 cm (side ward neutron beam direction) on theZ axis of target. On this position (60,30) counting result was 104358 for Pbcollimator and 246652 for PVC collimator. From examination result shown thatthe weight of silicon is in order 175 gram. (author)

  10. Shielding calculations for the SNO detector

    International Nuclear Information System (INIS)

    Earle, E.D.; Wong, P.Y.

    1987-05-01

    The gamma-ray background into the central D 2 O vessel of the SNO detector due to Th and U in the rock, concrete, and photomultipliers is calculated. A cylindrical geometry and concrete thicknesses of 0.5 and 1 m are assumed. The effect of adding boron to the concrete is also considered. It is concluded that backgrounds from (α,n) reactions can be reduced to the required level. These calculations will assist in finalizing the detector design but additional calculations will be required as new design details become known

  11. Detector Background Reduction by Passive and Active Shielding

    International Nuclear Information System (INIS)

    Bikit, I.; Bikit, K.; Forkapic, S.; Mrda, D.; Nikolov, J.; Slivka, J.; Todorovic, N.

    2013-01-01

    The operational problems of the gamma ray spectrometer shielded passively with 12 cm of lead and actively by five 0.5 m × 0.5 m × 0.05 m plastic veto shields are described. The active shielding effect from both environmental gamma ray, cosmic muons and neutrons was investigated. For anticoincidence gating wide range of scintillator pulses, corresponding to the energy range of 150 keV-75 MeV, were used. With the optimal set up the integral background, for the energy region of 50 - 3000 keV, of 0.31 c/s was achieved. The detector mass related background was 0.345 c/(kg s). The 511 keV annihilation line was reduced by the factor of 7 by the anticoincidence gate. It is shown that the plastic shields increase the neutron capture gamma line intensities due to neutron termalization.(author)

  12. Spectral perturbations from silicon diode detector encapsulation and shielding in photon fields.

    Science.gov (United States)

    Eklund, Karin; Ahnesjö, Anders

    2010-11-01

    Silicon diodes are widely used as detectors for relative dose measurements in radiotherapy. The common manufacturing practice is to encapsulate the diodes in plastic for protection and to facilitate mounting in scanning devices. Diodes intended for use in photon fields commonly also have a shield of a high atomic number material (usually tungsten) integrated into the encapsulation to selectively absorb low-energy photons to which silicon diodes would otherwise over-response. However, new response models based on cavity theories and spectra calculations have been proposed for direct correction of the readout from unshielded (e.g., "electron") diodes used in photon fields. This raises the question whether it is correct to assume that the spectrum in a water phantom at the location of the detector cavity is not perturbed by the detector encapsulation materials. The aim of this work is to investigate the spectral effects of typical encapsulations, including shielding, used for clinical diodes. The effects of detector encapsulation of an unshielded and a shielded commercial diode on the spectra at the detector cavity location are studied through Monte Carlo simulations with PENELOPE-2005. Variance reduction based on correlated sampling is applied to reduce the CPU time needed for the simulations. The use of correlated sampling is found to be efficient and to not introduce any significant bias to the results. Compared to reference spectra calculated in water, the encapsulation for an unshielded diode is demonstrated to not perturb the spectrum, while a tungsten shielded diode caused not only the desired decrease in low-energy scattered photons but also a large increase of the primary electron fluence. Measurements with a shielded diode in a 6 MV photon beam proved that the shielding does not completely remove the field-size dependence of the detector response caused by the over-response from low-energy photons. Response factors of a properly corrected unshielded diode

  13. Background components of Ge(Li) and GeHP-detectors in the passive shield

    International Nuclear Information System (INIS)

    Buraeva, E.A.; Davydov, M.G.; Zorina, L.V.; Stasov, V.V.

    2007-01-01

    The gamma-spectrometer Ge(Li)- and the extra pure Ge-detector background components in a specially designed passive shield were subjected to investigation in the land-based laboratory in 1996-2006. The measurement time period varied from 45 up to 240 hours. The detector background is caused by the radionuclides in the shield material, in the shield cells and in the detector materials. The prominence was given to the study of the revealed time dependence of 222 Rn daughter product background including '2 10 Pb 46.5 keV peak [ru

  14. Implementation of Surface Detector Option in SCALE SAS4 Shielding Module

    International Nuclear Information System (INIS)

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

    1999-01-01

    The Shielding Analysis Sequence No. 4 (SAS4) in the Standardized Cask Analysis and Licensing Evaluation System (SCALE) is designed to aid the novice user in the preparation of detailed three-dimensional models and radiation protection studies of transportation or storage packages containing spent fuel from a nuclear reactor facility. The underlying methodology in these analyses is the Monte Carlo particle-tracking approach as incorporated into the MORSE-SGC computer code. The use of these basic procedures is enhanced via the automatic generation of the biasing parameters in the SAS4 sequence, which dramatically increases the calculational efficiency of most standard shielding problems. Until recently the primary mechanism for dose estimates in SAS4 was the use of point detectors, which were effective for single-dose locations, but inefficient for quantification of dose-rate profiles. This paper describes the implementation of a new surface detector option for SAS4 with automatic discretization of the detector surface into multiple segments or subdetectors. Results from several sample problems are given and discussed

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  16. Test of thermal shields for early warning station detectors

    DEFF Research Database (Denmark)

    Petersen, Jesper

    1997-01-01

    The properties of thermal shields around NaI crystal scintillators for early warning stations have been checked in order to assure that external temperature variations cannot influence the stability of the measurements....

  17. Activation of concrete samples from the biological shield of the ASTRA reactor

    International Nuclear Information System (INIS)

    Smecka, F.

    2006-09-01

    Drill cores from the biological shield of the ASTRA reactor in Seibersdorf were taken and milled because of the different size of the Baryt crystals in the concrete in order to get homogenous samples. The powder samples were put into bore holes of a graphite block which was placed into the thermal column of the TRIGA Mark II reactor. The block was irradiated for 10 minutes at a reactor power of 25 kW. After one hour the dose rate was examined and the samples were ready for further save handling. The gamma spectrum was measured with a Ge detector and the results were compared with simulation data. (nevyjel)

  18. Laboratory tests on neutron shields for gamma-ray detectors in space

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  19. Remote sampling and analysis of highly radioactive samples in shielded boxes

    International Nuclear Information System (INIS)

    Kirpikov, D.A.; Miroshnichenko, I.V.; Pykhteev, O.Yu.

    2010-01-01

    The sampling procedure used for highly radioactive coolant water is associated with high risk of personnel irradiation and uncontrolled radioactive contamination. Remote sample manipulation with provision for proper radiation shielding is intended for safety enhancement of the sampling procedure. The sampling lines are located in an isolated compartment, a shielded box. Various equipment which enables remote or automatic sample manipulation is used for this purpose. The main issues of development of the shielded box equipment intended for a wider ranger of remote chemical analyses and manipulation techniques for highly radioactive water samples are considered in the paper. There were three principal directions of work: Transfer of chemical analysis performed in the laboratory inside the shielded box; Prevalence of computer-aided and remote techniques of highly radioactive sample manipulation inside the shielded box; and, Increase in control over sampling and determination of thermal-hydraulic parameters of the coolant water in the sampling lines. The developed equipment and solutions enable remote chemical analysis in the restricted volume of the shielded box by using ion-chromatographic, amperometrical, fluorimetric, flow injection, phototurbidimetric, conductometric and potentiometric methods. Extent of control performed in the shielded box is determined taking into account the requirements of the regulatory documents as well as feasibility and cost of the technical adaptation of various methods to the shielded box conditions. The work resulted in highly precise determination of more than 15 indexes of the coolant water quality performed in on-line mode in the shielded box. It averages to 80% of the total extent of control performed at the prototype reactor plants. The novel solutions for highly radioactive sample handling are implemented in the shielded box (for example, packaging, sample transportation to the laboratory, volume measurement). The shielded box is

  20. Optimizing moderation of He-3 neutron detectors for shielded fission sources

    Energy Technology Data Exchange (ETDEWEB)

    Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu [Department of Physics and Astronomy, Brigham Young University, Provo, UT 84602 (United States); Czirr, J. Bart, E-mail: czirr@juno.com [Department of Physics and Astronomy, Brigham Young University, Provo, UT 84602 (United States)

    2012-11-01

    The response of a {sup 3}He neutron detector is highly dependent on the amount of moderator incorporated into the detector system. If there is too little moderation, neutrons will not react with the {sup 3}He. If there is too much moderation, neutrons will not reach the {sup 3}He. In applications for portal or border monitors where {sup 3}He detectors are used to interdict illicit importation of plutonium, the fission source is always shielded to some extent. Since the energy distribution of neutrons emitted from the source depends on the amount and type of shielding present, the optimum placement of moderating material around {sup 3}He tubes is a function of shielding. In this paper, we use Monte Carlo techniques to model the response of {sup 3}He tubes placed in polyethylene boxes for moderation. To model the shielded fission neutron source, we use a point {sup 252}Cf source placed in the center of polyethylene spheres of varying radius. Detector efficiency as a function of box geometry and shielding is explored. We find that increasing the amount of moderator behind and to the sides of the detector generally improves the detector response, but that incremental benefits are minimal if the thickness of the polyethylene moderator is greater than about 5-7 cm. The thickness of the moderator in front of the {sup 3}He tubes, however, is very important. For bare sources, about 4-5 cm of moderator is optimum, but as the shielding increases, the optimum thickness of this moderator decreases to 0.5-1 cm. Similar conclusions can be applied to polyethylene boxes employing two {sup 3}He tubes. Two-tube boxes with front moderators of non-uniform thickness may be useful for detecting neutrons over a wide energy range.

  1. Determination of self shielding factors and gamma attenuation effects for tree ring samples

    International Nuclear Information System (INIS)

    Dagistan Sahin; Kenan Uenlue

    2012-01-01

    Determination of tree ring chemistry using Neutron Activation Analysis (NAA) is part of an ongoing research between Penn State University (PSU) and Cornell University, The Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology. Tree-ring chemistry yields valuable data for environmental event signatures. These signatures are a complex function of elemental concentration. To be certain about concentration of signature elements, it is necessary to perform the measurements and corrections with the lowest error and maximum accuracy possible. Accurate and precise values of energy dependent neutron flux at dry irradiation tubes and detector efficiency for tree ring sample are calculated for Penn State Breazeale Reactor (PSBR). For the calculation of energy dependent and self shielding corrected neutron flux, detailed model of the TRIGA Mark III reactor at PSU with updated fuel compositions was prepared using the MCNP utility for reactor evolution (MURE) libraries. Dry irradiation tube, sample holder and sample were also included in the model. The thermal flux self-shielding correction factors due to the sample holder and sample for were calculated and verified with previously published values. The Geant-4 model of the gamma spectroscopy system, developed at Radiation Science and Engineering Center (RSEC), was improved and absolute detector efficiency for tree-ring samples was calculated. (author)

  2. Monolithic active pixel radiation detector with shielding techniques

    Energy Technology Data Exchange (ETDEWEB)

    Deptuch, Grzegorz W.

    2018-03-20

    A monolithic active pixel radiation detector including a method of fabricating thereof. The disclosed radiation detector can include a substrate comprising a silicon layer upon which electronics are configured. A plurality of channels can be formed on the silicon layer, wherein the plurality of channels are connected to sources of signals located in a bulk part of the substrate, and wherein the signals flow through electrically conducting vias established in an isolation oxide on the substrate. One or more nested wells can be configured from the substrate, wherein the nested wells assist in collecting charge carriers released in interaction with radiation and wherein the nested wells further separate the electronics from the sensing portion of the detector substrate. The detector can also be configured according to a thick SOA method of fabrication.

  3. Sample Selection for Training Cascade Detectors.

    Science.gov (United States)

    Vállez, Noelia; Deniz, Oscar; Bueno, Gloria

    2015-01-01

    Automatic detection systems usually require large and representative training datasets in order to obtain good detection and false positive rates. Training datasets are such that the positive set has few samples and/or the negative set should represent anything except the object of interest. In this respect, the negative set typically contains orders of magnitude more images than the positive set. However, imbalanced training databases lead to biased classifiers. In this paper, we focus our attention on a negative sample selection method to properly balance the training data for cascade detectors. The method is based on the selection of the most informative false positive samples generated in one stage to feed the next stage. The results show that the proposed cascade detector with sample selection obtains on average better partial AUC and smaller standard deviation than the other compared cascade detectors.

  4. Sample Selection for Training Cascade Detectors.

    Directory of Open Access Journals (Sweden)

    Noelia Vállez

    Full Text Available Automatic detection systems usually require large and representative training datasets in order to obtain good detection and false positive rates. Training datasets are such that the positive set has few samples and/or the negative set should represent anything except the object of interest. In this respect, the negative set typically contains orders of magnitude more images than the positive set. However, imbalanced training databases lead to biased classifiers. In this paper, we focus our attention on a negative sample selection method to properly balance the training data for cascade detectors. The method is based on the selection of the most informative false positive samples generated in one stage to feed the next stage. The results show that the proposed cascade detector with sample selection obtains on average better partial AUC and smaller standard deviation than the other compared cascade detectors.

  5. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

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

    2007-01-01

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

  6. Method for optimizing side shielding in positron-emission tomographs and for comparing detector materials

    International Nuclear Information System (INIS)

    Derenzo, S.E.

    1980-01-01

    This report presents analytical formulas for the image-forming and background event rates seen by circular positron-emission tomographs with parallel side shielding. These formulas include deadtime losses, detector efficiency, coincidence resolving time, amount of activity, patient port diameter, shielding gap, and shielding depth. A figure of merit, defined in terms of these quantities, describes the signal-to-noise ratio in the reconstructed image of a 20-cm cylinder of water with uniformly dispersed activity. Results are presented for the scintillators NaI(TI), bismuth germanate (BGO), CsF, and plastic; and for Ge(Li) and wire chambers with converters. In these examples, BGO provided the best signal-to-noise for activity levels below 1000 μCi per cm, and CsF had the advantage for higher activity levels

  7. Induced radioactivity in the forward shielding and semiconductor tracker of the ATLAS detector.

    Science.gov (United States)

    Bĕdajánek, I; Linhart, V; Stekl, I; Pospísil, S; Kolros, A; Kovalenko, V

    2005-01-01

    The radioactivity induced in the forward shielding, copper collimator and semiconductor tracker modules of the ATLAS detector has been studied. The ATLAS detector is a long-term experiment which, during operation, will require to have service and access to all of its parts and components. The radioactivity induced in the forward shielding was calculated by Monte Carlo methods based on GEANT3 software tool. The results show that the equivalent dose rates on the outer surface of the forward shielding are very low (at most 0.038 microSv h(-1)). On the other hand, the equivalent dose rates are significantly higher on the inner surface of the forward shielding (up to 661 microSv h(-1)) and, especially, at the copper collimator close to the beampipe (up to 60 mSv h(-1)). The radioactivity induced in the semiconductor tracker modules was studied experimentally. The module was activated by neutrons in a training nuclear reactor and the delayed gamma ray spectra were measured. From these measurements, the equivalent dose rate on the surface of the semiconductor tracker module was estimated to be LHC) operation and 10 d of cooling.

  8. Monte Carlo simulations and measurements for efficiency determination of lead shielded plastic scintillator detectors

    Science.gov (United States)

    Yasin, Zafar; Negoita, Florin; Tabbassum, Sana; Borcea, Ruxandra; Kisyov, Stanimir

    2017-12-01

    The plastic scintillators are used in different areas of science and technology. One of the use of these scintillator detectors is as beam loss monitors (BLM) for new generation of high intensity heavy ion in superconducting linear accelerators. Operated in pulse counting mode with rather high thresholds and shielded by few centimeters of lead in order to cope with radiofrequency noise and X-ray background emitted by accelerator cavities, they preserve high efficiency for high energy gamma ray and neutrons produced in the nuclear reactions of lost beam particles with accelerator components. Efficiency calculation and calibration of detectors is very important before their practical usage. In the present work, the efficiency of plastic scintillator detectors is simulated using FLUKA for different gamma and neutron sources like, 60Co, 137Cs and 238Pu-Be. The sources are placed at different positions around the detector. Calculated values are compared with the measured values and a reasonable agreement is observed.

  9. Scintillation detector with anticoincidence shield for determination of the radioactive concentration of standard solutions

    International Nuclear Information System (INIS)

    Broda, R.; Radoszewski, T.

    1982-01-01

    The construction and parameters of the prototype liquid scintillation detector for disintegration rate determination of standard solutions is described. The detector is equipped with a liquid scintillation anticoincidence shield with a volume of 40 l. The instrument is placed in the building of the Radioisotope Production and Distribution Centre in the Institute of Nuclear Research at Swierk. The results of instrument background reduction are described. The counting efficiency of several beta-emitters 3 H, 63 Ni, 14 C and 90 Sr + 90 Y is given, as well as the examples of a disintegration rate determination of low radioactivity concentration of standard solutions. (author)

  10. Evaluation of neutron shielding properties of lead glass using bubble detector

    International Nuclear Information System (INIS)

    Viswanathan, S.; Vishwa Prasad, K.; Srinivasan, T.K.; Ponraju, D.

    1999-01-01

    Neutron shielding properties of lead glass had been studied using a 241 Am-Be neutron source. Indigenously developed bubble detector was used as neutron detector. Attenuation curves were determined experimentally for the lead glass under the conditions of broad beam geometry. Theoretical calculations were made using Monte Carlo code MCNP3. Measurements were made for polyethylene and concrete to serve as reference. The measured and calculated neutron removal cross sections of lead glass, polyethylene and concrete are reported in this paper. Good agreement is observed between the experimental results and theoretical calculations. (author)

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

    Directory of Open Access Journals (Sweden)

    Fernandes Ana C.

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    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.

  13. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    International Nuclear Information System (INIS)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-01-01

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼1.4e9/cm 2 . Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm 2 . The radiation damage and operational upsets which starts at ∼1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor ∼50

  14. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-11-02

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.

  15. Two specialized delayed-neutron detector designs for assays of fissionable elements in water and sediment samples

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Balagna, J.P.; Menlove, H.O.

    1976-01-01

    Two specialized neutron-sensitive detectors are described which are employed for rapid assays of fissionable elements by sensing for delayed neutrons emitted by samples after they have been irradiated in a nuclear reactor. The more sensitive of the two detectors, designed to assay for uranium in water samples, is 40% efficient; the other, designed for sediment sample assays, is 27% efficient. These detectors are also designed to operate under water as an inexpensive shielding against neutron leakage from the reactor and neutrons from cosmic rays. (Auth.)

  16. The GSF anticoincidence-shielded Ge(Li) gamma-ray spectrometer and its application to the analysis of environmental samples

    International Nuclear Information System (INIS)

    Hoetzl, H.; Winkler, R.

    1981-01-01

    A high-efficiency gamma-ray spectrometer has been designed and built to provide simultaneous anticoincidence and coincidence spectrometry of low-level environmental samples. The spectrometer consists of a large-volume Ge(Li) detector as the main detector and a well-type NaI(Tl) guard detector. The Ge(Li) detector is a closed-end coaxial detector housed in a crystal of the vertical dip-stick type. Its relative photopeak efficiency is 27.5%. The guard counter is a 23-cm-dia. by 23-cm-long NaI(Tl) crystal with a 7.8-cm-dia. by 18-cm-deep centre well. The passive shield consists of a 10-cm lead shield with copper and cadmium lining. The electronics is designed to operate independently and simultaneously in the anticoincidence mode as well as in the coincidence or in the normal passive shield mode. When operating in the anticoincidence mode the Compton edge of 137 Cs is reduced by a factor of 7.7 to provide a peak-to-Compton edge ratio of 480:1. Bulk samples up to about 300 cm 3 can be measured on the top of the detector end cap inside the well of the NaI(Tl) crystal. The lower limit of detection (1000 min counting time, 95% confidence level) for 137 Cs is 1.6 pCi in a 3.8-cm-dia. by 3.5-cm-high sample geometry. The design of the spectrometer, its properties and the application to investigations on the migration of radionuclides in the soil, the analysis of radioactive emissions of coal-fired power plants and to fallout studies are described. (author)

  17. Attenuation of a non-parallel beam of gamma radiation by thick shielding-application to the determination of the 235U enrichment with NaI detectors

    International Nuclear Information System (INIS)

    Mortreau, Patricia; Berndt, Reinhard

    2005-01-01

    The traditional method used to determine the Uranium enrichment by nondestructive analysis is based on the 'enrichment meter principle' [1]. It involves measuring the intensity of the 186 keV net peak area of 235 U in 'quasi-infinite' samples. A prominent factor, which affects the peak intensity, is the presence of gamma absorbing material (e.g., container wall, detector cover) between the sample and the detector. Its effect is taken into consideration in a commonly called 'wall thickness' correction factor. Often calculated on the basis of approximations, its performance is adequate for small attenuation factors applicable to the case of narrow beams. However these approximations do not lead to precise results when wide non-parallel beams are attenuated through thick container walls. This paper is dedicated to the calculation by numerical integration of the geometrical correction factor (K wtc ) which describes the effective mean path length of the radiation through the absorbing layer. This factor was calculated as a function of various measurement parameters (types and dimensions of the detector, of the collimator and of the shielding) for the most commonly used collimator shapes and detectors. Both coherent scattering (Rayleigh) and incoherent scattering (Compton) are taken into account for the calculation of the radiation interaction within the detector

  18. Results of substitution of the Nal by a Ge detector in a simple shadow shield whole body counter

    International Nuclear Information System (INIS)

    Sahre, P.; Schoenmuth, T.; Thieme, K.

    1997-01-01

    Since 1976 a whole body counter (WBC) has been used at the Rossendorf Research Centre for measuring the internal contamination of workers. The WBC with the Germanium detector is given schematically and visually. The WBC is a shadow shield type with a tilted chair having only one detector. Table 1 contains the parameters of the WBC. It can be seen that the WBC is a simple counter. Therefore, taking into account the experiences of McCurdy, a lot of improvements were expected form the simple substitution of a HP Germanium detector for a NaI (TI) detector, i.e. despite a decrease in the sensitive detection volume, an enhancement of all quantifiable results (e.g. lower limit of detection and time for analysis of the spectrum) and above all the reliability and automation of nuclide identification were expected. (orig./SR)

  19. Results of substitution of the Nal by a Ge detector in a simple shadow shield whole body counter

    Energy Technology Data Exchange (ETDEWEB)

    Sahre, P.; Schoenmuth, T. [Nuclear Engineering and Analytics Inc. Rossendorf, Dresden (Germany); Thieme, K. [Amersham Buchler Ltd. und Co., Braunschweig (Germany)

    1997-12-01

    Since 1976 a whole body counter (WBC) has been used at the Rossendorf Research Centre for measuring the internal contamination of workers. The WBC with the Germanium detector is given schematically and visually. The WBC is a shadow shield type with a tilted chair having only one detector. Table 1 contains the parameters of the WBC. It can be seen that the WBC is a simple counter. Therefore, taking into account the experiences of McCurdy, a lot of improvements were expected form the simple substitution of a HP Germanium detector for a NaI (TI) detector, i.e. despite a decrease in the sensitive detection volume, an enhancement of all quantifiable results (e.g. lower limit of detection and time for analysis of the spectrum) and above all the reliability and automation of nuclide identification were expected. (orig./SR)

  20. Fractionation of plutonium in environmental and bio-shielding concrete samples using dynamic sequential extraction

    DEFF Research Database (Denmark)

    Qiao, Jixin; Hou, Xiaolin

    2010-01-01

    Fractionation of plutonium isotopes (238Pu, 239,240Pu) in environmental samples (i.e. soil and sediment) and bio-shielding concrete from decommissioning of nuclear reactor were carried out by dynamic sequential extraction using an on-line sequential injection (SI) system combined with a specially...

  1. Shielding design of highly activated sample storage at reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Naim Syauqi Hamzah; Julia Abdul Karim; Mohamad Hairie Rabir; Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh

    2010-01-01

    Radiation protection has always been one of the most important things considered in Reaktor Triga PUSPATI (RTP) management. Currently, demands on sample activation were increased from variety of applicant in different research field area. Radiological hazard may occur if the samples evaluation done were misjudge or miscalculated. At present, there is no appropriate storage for highly activated samples. For that purpose, special irradiated samples storage box should be provided in order to segregate highly activated samples that produce high dose level and typical activated samples that produce lower dose level (1 - 2 mR/ hr). In this study, thickness required by common shielding material such as lead and concrete to reduce highly activated radiotracer sample (potassium bromide) with initial exposure dose of 5 R/ hr to background level (0.05 mR/ hr) were determined. Analyses were done using several methods including conventional shielding equation, half value layer calculation and Micro shield computer code. Design of new irradiated samples storage box for RTP that capable to contain high level gamma radioactivity were then proposed. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  3. Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

    International Nuclear Information System (INIS)

    Noorddin Ibrahim; Rosnie Akang

    2009-01-01

    Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

  4. Sample Selection for Training Cascade Detectors

    OpenAIRE

    V?llez, Noelia; Deniz, Oscar; Bueno, Gloria

    2015-01-01

    Automatic detection systems usually require large and representative training datasets in order to obtain good detection and false positive rates. Training datasets are such that the positive set has few samples and/or the negative set should represent anything except the object of interest. In this respect, the negative set typically contains orders of magnitude more images than the positive set. However, imbalanced training databases lead to biased classifiers. In this paper, we focus our a...

  5. Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

    Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

  6. Evaluation of the performance of the shields in the EPMAs used for radioactive samples

    International Nuclear Information System (INIS)

    Matsui, Hiroki; Suzuki, Miho; Obata, Hiroki; Kanazawa, Hiroyuki

    2014-06-01

    The Reactor Fuel Examination Facility (RFEF) in Japan Atomic Energy Agency (JAEA) has been used for Post Irradiation Examinations (PIEs) to verify the reliability and safety of the nuclear fuels irradiated in commercial reactors. EPMA (Electron Probe Micro Analyzer) has been utilized for the qualitative analysis of the fission product in the fuel pellet and the detailed observation of the oxide layers formed at the inner and outer surfaces of fuel cladding. Commercial EPMAs were remodeled so that the EPMAs can be applied for radioactive samples. Several shields were set in the EPMA to avoid the gamma-rays which radiate from a radioactive sample to the proportional counter in the EPMA. It is important to calculate this shielding performance adequately to maintain the precision of analysis. This report describes the results of re-evaluation of the performance of the shields in the EPMAs in the RFEF by using the Particle and Heavy Ion Transport Code System (PHITS) and the examination results of gamma-ray effect to the X-ray spectrum data by using a radioactive sample. (author)

  7. Design and Characterization of a Gradient-Transparent RF Copper Shield for PET Detector Modules in Hybrid MR-PET Imaging

    Science.gov (United States)

    Berneking, Arne; Trinchero, Riccardo; Ha, YongHyun; Finster, Felix; Cerello, Piergiorgio; Lerche, Christoph; Shah, Nadim Jon

    2017-05-01

    This paper focuses on the design and the characterization of a frequency-selective shield for positron emission tomography (PET) detector modules of hybrid magnetic resonance-PET scanners, where the shielding of the PET cassettes is located close to the observed object. The proposed shielding configuration is designed and optimized to guarantee a high shielding effectiveness (SE) of up to 60 dB for B1-fields at the Larmor frequency of 64 MHz, thus preventing interactions between the radio-frequency (RF) coil and PET electronics. On the other hand, the shield is transparent to the gradient fields with the consequence that eddy-current artifacts in the acquired EPI images are significantly reduced with respect to the standard solid-shield configuration. The frequency-selective behavior of the shield is characterized and validated via simulation studies with CST MICROWAVE STUDIO in the megahertz and kilohertz range. Bench measurements with an RF coil built in-house demonstrated the high SE at the Larmor frequency. Moreover, measurements on a 4-T human scanner confirmed the abolishment of eddy current artifact and also provided an understanding of where the eddy currents occur with respect to the sequence parameters. Simulations and measurements for the proposed shielding concept were compared with a solid copper shielding configuration.

  8. A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Thanh, Tran Thien; Tao, Chau Van; Loan, Truong Thi Hong; Nhon, Mai Van; Chuong, Huynh Dinh; Au, Bui Hai [Vietnam National Univ., Ho Chi Minh City (Viet Nam). Dept. of Nuclear Physics

    2012-12-15

    The accuracy of the coincidence-summing corrections in gamma spectrometry depends on the total efficiency calibration that is hardly obtained over the whole energy as the required experimental conditions are not easily attained. Monte Carlo simulations using MCNP5 code was performed in order to estimate the affect of the shielding to total efficiency. The effect of HPGe response are also shown. (orig.)

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

    International Nuclear Information System (INIS)

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

    1998-09-01

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

  10. Equipment and detectors calibration behind shielding of CERN high-energy particle accelerator SPS: June 2007

    International Nuclear Information System (INIS)

    Spurny, Frantisek; Ploc, Ondrej

    2008-02-01

    High energy stray radiation fields have been realised since 1993 at CERN also in the frame of CEC-CERN collaboration on the project: 'Detection and the Dosimetry of Neutrons and Charged Particles at Aviation Altitudes in the Earth's Atmosphere'. They are formed at the H6 beam of the north experimental area of the SPS facility. Two shielding configurations have been built, with the top concrete, resp. top iron shielding. Many intercalibration experiments have been realised since the beginning. After an interruption due to technical problems, two other campaigns have been realised during 2006 year, another one during the June 2007. This report describes analyses and discusses the most of results obtained during the last, 2007 run.. (author)

  11. The shielding properties of the newly developed container for transport of samples contaminated with CBRN substances

    International Nuclear Information System (INIS)

    Fisera, O.; Kares, J.

    2014-01-01

    A container for transport of environmental samples to the analytical laboratory is being developed as part of the development of system for collection and transport of samples contaminated with chemical, biological, radioactive and nuclear (CBRN) substances after CBRN incidents. The proposed system corresponds with current requirements of NATO publication AEP-66. The proposed container will meet the requirements of mechanical stability and tightness for the packaging of the chemical, biological and radioactive substances. Verification of shielding properties and satisfaction of requirements of radiation protection during transport of potentially relatively high active samples was the aim of this part of research. The results, together with a wall thickness of the inner steel container, the inner lining and the outer transport package, give excellent assumption that the radiation protection requirements for the proposed container and transport package will be satisfied. (authors)

  12. A phoswich detector design for improved spatial sampling in PET

    Science.gov (United States)

    Thiessen, Jonathan D.; Koschan, Merry A.; Melcher, Charles L.; Meng, Fang; Schellenberg, Graham; Goertzen, Andrew L.

    2018-02-01

    Block detector designs, utilizing a pixelated scintillator array coupled to a photosensor array in a light-sharing design, are commonly used for positron emission tomography (PET) imaging applications. In practice, the spatial sampling of these designs is limited by the crystal pitch, which must be large enough for individual crystals to be resolved in the detector flood image. Replacing the conventional 2D scintillator array with an array of phoswich elements, each consisting of an optically coupled side-by-side scintillator pair, may improve spatial sampling in one direction of the array without requiring resolving smaller crystal elements. To test the feasibility of this design, a 4 × 4 phoswich array was constructed, with each phoswich element consisting of two optically coupled, 3 . 17 × 1 . 58 × 10mm3 LSO crystals co-doped with cerium and calcium. The amount of calcium doping was varied to create a 'fast' LSO crystal with decay time of 32.9 ns and a 'slow' LSO crystal with decay time of 41.2 ns. Using a Hamamatsu R8900U-00-C12 position-sensitive photomultiplier tube (PS-PMT) and a CAEN V1720 250 MS/s waveform digitizer, we were able to show effective discrimination of the fast and slow LSO crystals in the phoswich array. Although a side-by-side phoswich array is feasible, reflections at the crystal boundary due to a mismatch between the refractive index of the optical adhesive (n = 1 . 5) and LSO (n = 1 . 82) caused it to behave optically as an 8 × 4 array rather than a 4 × 4 array. Direct coupling of each phoswich element to individual photodetector elements may be necessary with the current phoswich array design. Alternatively, in order to implement this phoswich design with a conventional light sharing PET block detector, a high refractive index optical adhesive is necessary to closely match the refractive index of LSO.

  13. Optimisation of beam-pipe shielding for MUCH detector of CBM experiment

    International Nuclear Information System (INIS)

    Ahmad, S.; Farooq, M.; Chattopadhyay, S.

    2014-01-01

    The Compressed Baryonic Matter (CBM) experiment is one of the major scientific pillars of FAIR. The main goal of the experiment is to explore the Quantum Chromodynamics (QCD) phase diagram in the regions of high baryonic densities and moderate temperatures in the beam energy range of 10-45 AGeV. This includes also the search for the critical point, the first order deconfinement phase transition from the hadronic matter to the partonic matter and the study of equation-of-state of dense baryonic matter. The CBM research program comprises a comprehensive scan of observables, beam energies and collision systems. The observables includes low-mass dilepton pairs, charmonia and open charm, but also collective flow of rare and bulk particles, correlations and fluctuations. Some of the particles under study have low cross-sections (like charm) or small branching ratios (like low-mass vector mesons). Therefore, in order to compensate for the low yield the measurements have to be performed at very high reaction rates of up to about 10 MHz. These conditions demand for fast and radiation hard detectors and associated fast electronics, readout and online event reconstruction. Low material budget is required with in the detector acceptance to avoid multiple scattering which would limit high-precision measurements

  14. Attenuation of a non-parallel beam of gamma radiation by thick shielding-application to the determination of the {sup 235}U enrichment with NaI detectors

    Energy Technology Data Exchange (ETDEWEB)

    Mortreau, Patricia [European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800 Via Fermi, Ispra (Vatican City State, Holy See,) (Italy)]. E-mail: patricia.mortreau@jrc.it; Berndt, Reinhard [European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800 Via Fermi, Ispra (VA) (Italy)

    2005-09-21

    The traditional method used to determine the Uranium enrichment by nondestructive analysis is based on the 'enrichment meter principle' [1]. It involves measuring the intensity of the 186 keV net peak area of {sup 235}U in 'quasi-infinite' samples. A prominent factor, which affects the peak intensity, is the presence of gamma absorbing material (e.g., container wall, detector cover) between the sample and the detector. Its effect is taken into consideration in a commonly called 'wall thickness' correction factor. Often calculated on the basis of approximations, its performance is adequate for small attenuation factors applicable to the case of narrow beams. However these approximations do not lead to precise results when wide non-parallel beams are attenuated through thick container walls. This paper is dedicated to the calculation by numerical integration of the geometrical correction factor (K {sub wtc}) which describes the effective mean path length of the radiation through the absorbing layer. This factor was calculated as a function of various measurement parameters (types and dimensions of the detector, of the collimator and of the shielding) for the most commonly used collimator shapes and detectors. Both coherent scattering (Rayleigh) and incoherent scattering (Compton) are taken into account for the calculation of the radiation interaction within the detector.

  15. CdTe detector based PIXE mapping of geological samples

    Energy Technology Data Exchange (ETDEWEB)

    Chaves, P.C., E-mail: cchaves@ctn.ist.utl.pt [Centro de Física Atómica da Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal); IST/ITN, Instituto Superior Técnico, Universidade Técnica de Lisboa, Campus Tecnológico e Nuclear, EN10, 2686-953 Sacavém (Portugal); Taborda, A. [Centro de Física Atómica da Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal); IST/ITN, Instituto Superior Técnico, Universidade Técnica de Lisboa, Campus Tecnológico e Nuclear, EN10, 2686-953 Sacavém (Portugal); Oliveira, D.P.S. de [Laboratório Nacional de Energia e Geologia (LNEG), Apartado 7586, 2611-901 Alfragide (Portugal); Reis, M.A. [Centro de Física Atómica da Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal); IST/ITN, Instituto Superior Técnico, Universidade Técnica de Lisboa, Campus Tecnológico e Nuclear, EN10, 2686-953 Sacavém (Portugal)

    2014-01-01

    A sample collected from a borehole drilled approximately 10 km ESE of Bragança, Trás-os-Montes, was analysed by standard and high energy PIXE at both CTN (previous ITN) PIXE setups. The sample is a fine-grained metapyroxenite grading to coarse-grained in the base with disseminated sulphides and fine veinlets of pyrrhotite and pyrite. Matrix composition was obtained at the standard PIXE setup using a 1.25 MeV H{sup +} beam at three different spots. Medium and high Z elemental concentrations were then determined using the DT2fit and DT2simul codes (Reis et al., 2008, 2013 [1,2]), on the spectra obtained in the High Resolution and High Energy (HRHE)-PIXE setup (Chaves et al., 2013 [3]) by irradiation of the sample with a 3.8 MeV proton beam provided by the CTN 3 MV Tandetron accelerator. In this paper we present results, discuss detection limits of the method and the added value of the use of the CdTe detector in this context.

  16. Helical tomotherapy shielding calculation for an existing LINAC treatment room: sample calculation and cautions

    International Nuclear Information System (INIS)

    Wu Chuan; Guo Fanqing; Purdy, James A

    2006-01-01

    This paper reports a step-by-step shielding calculation recipe for a helical tomotherapy unit (TomoTherapy Inc., Madison, WI, USA), recently installed in an existing Varian 600C treatment room. Both primary and secondary radiations (leakage and scatter) are explicitly considered. A typical patient load is assumed. Use factor is calculated based on an analytical formula derived from the tomotherapy rotational beam delivery geometry. Leakage and scatter are included in the calculation based on corresponding measurement data as documented by TomoTherapy Inc. Our calculation result shows that, except for a small area by the therapists' console, most of the existing Varian 600C shielding is sufficient for the new tomotherapy unit. This work cautions other institutions facing the similar situation, where an HT unit is considered for an existing LINAC treatment room, more secondary shielding might be considered at some locations, due to the significantly increased secondary shielding requirement by HT. (note)

  17. Onboard cross-calibration of the Pille-ISS Detector System and measurement of radiation shielding effect of the water filled protective curtain in the ISS crew cabin

    International Nuclear Information System (INIS)

    Szántó, P.; Apáthy, I.; Deme, S.; Hirn, A.; Nikolaev, I.V.; Pázmándi, T.; Shurshakov, V.A.; Tolochek, R.V.; Yarmanova, E.N.

    2015-01-01

    As a preparation for long duration space missions it is important to determine and minimize the impact of space radiation on human health. One of the methods to diminish the radiation burden is using an additional local shielding in the places where the crewmembers can stay for longer time. To increase the crew cabin shielding a special protective curtain was designed and delivered to ISS in 2010 containing four layers of hygienic wipes and towels providing an additional shielding thickness of about 8 g/cm"2 water-equivalent matter. The radiation shielding effect of the protective curtain, in terms of absorbed dose, was measured with the thermoluminescent Pille-ISS Detector System. In order to verify the reliability of the Pille system an onboard cross-calibration was also performed. The measurement proved that potentially 25% reduction of the absorbed dose rate in the crew cabin can be achieved, that results in 8% (∼16 μGy/day) decrease of the total absorbed dose to the crew, assuming that they spend 8 h in the crew cabin a day. - Highlights: • The dose level in the ISS Zvezda crew quarters is higher than the average dose level in the module. • A shielding made of hygienic wipes and towels was set up onboard as additional protection. • Onboard cross calibration of the Pille-ISS space dosimeter (TL) system was performed. • The shielding effect of the protective curtain in terms of absorbed dose was measured with the onboard Pille system. • The shielding effect of the protective water curtain is approximately 24 ± 9% in absorbed dose.

  18. Radiation shielding lead shield

    International Nuclear Information System (INIS)

    Dei, Shoichi.

    1991-01-01

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

  19. Effect of physical, chemical and electro-kinetic properties of pumice samples on radiation shielding properties of pumice material

    International Nuclear Information System (INIS)

    Tapan, Mücip; Yalçın, Zeynel; İçelli, Orhan; Kara, Hüsnü; Orak, Salim; Özvan, Ali; Depci, Tolga

    2014-01-01

    Highlights: • Radiation shielding properties of pumice materials are studied. • The relationship between physical, chemical and electro-kinetic properties pumice samples is identified. • The photon atomic parameters are important for the absorber peculiarity of the pumices. - Abstract: Pumice has been used in cement, concrete, brick, and ceramic industries as an additive and aggregate material. In this study, some gamma-ray photon absorption parameters such as the total mass attenuation coefficients, effective atomic number and electronic density have been investigated for six different pumice samples. Numerous values of energy related parameters from low energy (1 keV) to high energy (100 MeV) were calculated using WinXCom programme. The relationship between radiation shielding properties of the pumice samples and their physical, chemical and electro-kinetic properties was evaluated using simple regression analysis. Simple regression analysis indicated a strong correlation between photon energy absorption parameters and density and SiO 2 , Fe 2 O 3 , CaO, MgO, TiO 2 content of pumice samples in this study. It is found that photon energy absorption parameters are not related to electro-kinetic properties of pumice samples

  20. A Detector System for Identifying Substances in a Sample

    DEFF Research Database (Denmark)

    2010-01-01

    for illuminating a first area and a second area of the corresponding cantilever (330); a detector array (334, 336), and a receiver diffractive optical element (329) adapted to collect light reflected from the cantilever array and to form at least two spatially overlapping images of said areas so that interference...

  1. Characteristic Performance Evaluation of a new SAGe Well Detector for Small and Large Sample Geometries

    International Nuclear Information System (INIS)

    Adekola, A.S.; Colaresi, J.; Douwen, J.; Jaederstroem, H.; Mueller, W.F.; Yocum, K.M.; Carmichael, K.

    2015-01-01

    Environmental scientific research requires a detector that has sensitivity low enough to reveal the presence of any contaminant in the sample at a reasonable counting time. Canberra developed the germanium detector geometry called Small Anode Germanium (SAGe) Well detector, which is now available commercially. The SAGe Well detector is a new type of low capacitance germanium well detector manufactured using small anode technology capable of advancing many environmental scientific research applications. The performance of this detector has been evaluated for a range of sample sizes and geometries counted inside the well, and on the end cap of the detector. The detector has energy resolution performance similar to semi-planar detectors, and offers significant improvement over the existing coaxial and Well detectors. Energy resolution performance of 750 eV Full Width at Half Maximum (FWHM) at 122 keV γ-ray energy and resolution of 2.0 - 2.3 keV FWHM at 1332 keV γ-ray energy are guaranteed for detector volumes up to 425 cm 3 . The SAGe Well detector offers an optional 28 mm well diameter with the same energy resolution as the standard 16 mm well. Such outstanding resolution performance will benefit environmental applications in revealing the detailed radionuclide content of samples, particularly at low energy, and will enhance the detection sensitivity resulting in reduced counting time. The detector is compatible with electric coolers without any sacrifice in performance and supports the Canberra Mathematical efficiency calibration method (In situ Object Calibration Software or ISOCS, and Laboratory Source-less Calibration Software or LABSOCS). In addition, the SAGe Well detector supports true coincidence summing available in the ISOCS/LABSOCS framework. The improved resolution performance greatly enhances detection sensitivity of this new detector for a range of sample sizes and geometries counted inside the well. This results in lower minimum detectable

  2. Characteristic Performance Evaluation of a new SAGe Well Detector for Small and Large Sample Geometries

    Energy Technology Data Exchange (ETDEWEB)

    Adekola, A.S.; Colaresi, J.; Douwen, J.; Jaederstroem, H.; Mueller, W.F.; Yocum, K.M.; Carmichael, K. [Canberra Industries Inc., 800 Research Parkway, Meriden, CT 06450 (United States)

    2015-07-01

    Environmental scientific research requires a detector that has sensitivity low enough to reveal the presence of any contaminant in the sample at a reasonable counting time. Canberra developed the germanium detector geometry called Small Anode Germanium (SAGe) Well detector, which is now available commercially. The SAGe Well detector is a new type of low capacitance germanium well detector manufactured using small anode technology capable of advancing many environmental scientific research applications. The performance of this detector has been evaluated for a range of sample sizes and geometries counted inside the well, and on the end cap of the detector. The detector has energy resolution performance similar to semi-planar detectors, and offers significant improvement over the existing coaxial and Well detectors. Energy resolution performance of 750 eV Full Width at Half Maximum (FWHM) at 122 keV γ-ray energy and resolution of 2.0 - 2.3 keV FWHM at 1332 keV γ-ray energy are guaranteed for detector volumes up to 425 cm{sup 3}. The SAGe Well detector offers an optional 28 mm well diameter with the same energy resolution as the standard 16 mm well. Such outstanding resolution performance will benefit environmental applications in revealing the detailed radionuclide content of samples, particularly at low energy, and will enhance the detection sensitivity resulting in reduced counting time. The detector is compatible with electric coolers without any sacrifice in performance and supports the Canberra Mathematical efficiency calibration method (In situ Object Calibration Software or ISOCS, and Laboratory Source-less Calibration Software or LABSOCS). In addition, the SAGe Well detector supports true coincidence summing available in the ISOCS/LABSOCS framework. The improved resolution performance greatly enhances detection sensitivity of this new detector for a range of sample sizes and geometries counted inside the well. This results in lower minimum detectable

  3. An improved correlated sampling method for calculating correction factor of detector

    International Nuclear Information System (INIS)

    Wu Zhen; Li Junli; Cheng Jianping

    2006-01-01

    In the case of a small size detector lying inside a bulk of medium, there are two problems in the correction factors calculation of the detectors. One is that the detector is too small for the particles to arrive at and collide in; the other is that the ratio of two quantities is not accurate enough. The method discussed in this paper, which combines correlated sampling with modified particle collision auto-importance sampling, and has been realized on the MCNP-4C platform, can solve these two problems. Besides, other 3 variance reduction techniques are also combined with correlated sampling respectively to calculate a simple calculating model of the correction factors of detectors. The results prove that, although all the variance reduction techniques combined with correlated sampling can improve the calculating efficiency, the method combining the modified particle collision auto-importance sampling with the correlated sampling is the most efficient one. (authors)

  4. Laser solid sampling for a solid-state-detector ICP emission spectrometer

    International Nuclear Information System (INIS)

    Noelte, J.; Moenke-Blankenburg, L.; Schumann, T.

    1994-01-01

    Solid sampling with laser vaporization has been coupled to an ICP emission spectrometer with an Echelle optical system and a solid-state-detector for the analysis of steel and soil samples. Pulsation of the vaporized material flow was compensated by real-time background correction and internal standardization, resulting in good accuracy and precision. (orig.)

  5. Performance evaluation of continuous blood sampling system for PET study. Comparison of three detector-systems

    CERN Document Server

    Matsumoto, K; Sakamoto, S; Senda, M; Yamamoto, S; Tarutani, K; Minato, K

    2002-01-01

    To measure cerebral blood flow with sup 1 sup 5 O PET, it is necessary to measure the time course of arterial blood radioactivity. We examined the performance of three different types of continuous blood sampling system. Three kinds of continuous blood sampling system were used: a plastic scintillator-based beta detector (conventional beta detector (BETA)), a bismuth germinate (BGO)-based coincidence gamma detector (Pico-count flow-through detector (COINC)) and a Phoswich detector (PD) composed by a combination of plastic scintillator and BGO scintillator. Performance of these systems was evaluated for absolute sensitivity, count rate characteristic, sensitivity to background gamnra photons, and reproducibility for nylon tube geometry. The absolute sensitivity of the PD was 0.21 cps/Bq for sup 6 sup 8 Ga positrons at the center of the detector. This was approximately three times higher than BETA, two times higher than COINC. The value measured with BETA was stable, even when background radioactivity was incre...

  6. Calibration of diffusion barrier charcoal detectors and application to radon sampling in dwellings

    Energy Technology Data Exchange (ETDEWEB)

    Montero C, M.E.; Colmenero S, L.; Villalba, L.; Saenz P, J.; Cano J, A.; Moreno B, A.; Renteria V, M.; Herrera P, E.F. [Cento de Investigacion en Materiales Avanzados, S. C. Miguel de Cervantes 120, Complejo Industrial Chihuahua, Chihuahua, (Mexico); Cruz G, S. De la [Facultad de Enfermeria y Nutriologia, Universidad Autonoma de Chihuahua, Av. Politecnico Nacional 2714, Chihuahua, (Mexico); Lopez M, A. [Instituto Nacional de Investigaciones Nucleares, Apartado Postal 18-1027, 11801 Mexico D.F. (Mexico)

    2003-07-01

    Some calibration conditions of diffusion barrier charcoal canister (DBCC) detectors for measuring radon concentration in air were studied. A series of functional expressions and graphs were developed to describe relationship between radon concentration in air and the activity adsorbed in DBCC, when placed in small chambers. A semi-empirical expression for the DBCC calibration was obtained, based on the detector integration time and the adsorption coefficient of radon on activated charcoal. Both, the integration time for 10 % of DBCC of the same batch, and the adsorption coefficient of radon for the activated charcoal used in these detectors, were experimentally determined. Using these values as the calibration parameters, a semi-empirical calibration function was used for the interpretation of the radon activities in the detectors used for sampling more than 200 dwellings in the major cities of the state of Chihuahua, Mexico. (Author)

  7. Calibration of diffusion barrier charcoal detectors and application to radon sampling in dwellings

    International Nuclear Information System (INIS)

    Montero C, M.E.; Colmenero S, L.; Villalba, L.; Saenz P, J.; Cano J, A.; Moreno B, A.; Renteria V, M.; Herrera P, E.F.; Cruz G, S. De la; Lopez M, A.

    2003-01-01

    Some calibration conditions of diffusion barrier charcoal canister (DBCC) detectors for measuring radon concentration in air were studied. A series of functional expressions and graphs were developed to describe relationship between radon concentration in air and the activity adsorbed in DBCC, when placed in small chambers. A semi-empirical expression for the DBCC calibration was obtained, based on the detector integration time and the adsorption coefficient of radon on activated charcoal. Both, the integration time for 10 % of DBCC of the same batch, and the adsorption coefficient of radon for the activated charcoal used in these detectors, were experimentally determined. Using these values as the calibration parameters, a semi-empirical calibration function was used for the interpretation of the radon activities in the detectors used for sampling more than 200 dwellings in the major cities of the state of Chihuahua, Mexico. (Author)

  8. Evaluation of high performance data acquisition boards for simultaneous sampling of fast signals from PET detectors

    International Nuclear Information System (INIS)

    Judenhofer, Martin S; Pichler, Bernd J; Cherry, Simon R

    2005-01-01

    Detectors used for positron emission tomography (PET) provide fast, randomly distributed signals that need to be digitized for further processing. One possibility is to sample the signals at the peak initiated by a trigger from a constant fraction discriminator (CFD). For PET detectors, simultaneous acquisition of many channels is often important. To develop and evaluate novel PET detectors, a flexible, relatively low cost and high performance laboratory data acquisition (DAQ) system is therefore required. The use of dedicated DAQ systems, such as a multi-channel analysers (MCAs) or continuous sampling boards at high rates, is expensive. This work evaluates the suitability of well-priced peripheral component interconnect (PCI)-based 8-channel DAQ boards (PD2-MFS-8 2M/14 and PD2-MFS-8-500k/14, United Electronic Industries Inc., Canton, MA, USA) for signal acquisition from novel PET detectors. A software package was developed to access the board, measure basic board parameters, and to acquire, visualize, and analyse energy spectra and position profiles from block detectors. The performance tests showed that the boards input linearity is >99.2% and the standard deviation is 22 Na source was 14.9% (FWHM) at 511 keV and is slightly better than the result obtained with a high-end single channel MCA (8000A, Amptek, USA) using the same detector (16.8%). The crystals (1.2 x 1.2 x 12 mm 3 ) within a 9 x 9 LSO block detector could be clearly separated in an acquired position profile. Thus, these boards are well suited for data acquisition with novel detectors developed for nuclear imaging

  9. Elevator mechanism and method for scintillation detectors

    International Nuclear Information System (INIS)

    Frank, E.

    1975-01-01

    An elevator mechanism and method for raising and lowering radioactive samples through a shielded vertical counting chamber in a benchtop scintillation detector is described. The elevator mechanism adds little or nothing to the height of the detector by using an elongated flexible member such as a metal tape secured to the bottom of the elevator platform and extending downwardly through the counting chamber and its bottom shielding, where the tape is bent laterally for connection to a drive means. In the particular embodiment illustrated, the tape is bent laterally below the bottom shielding for the counting chamber, and then upwardly along or through one side of the shielding to a reel at the top of the shielding. The tape is wound onto the reel, and the reel is driven by a reversible motor which winds and unwinds the tape on the reel to raise and lower the elevator platform

  10. Method of quantitative analysis of fluorine in environmental samples using a pure-Ge detector

    International Nuclear Information System (INIS)

    Sera, K.; Terasaki, K.; Saitoh, Y.; Itoh, J.; Futatsugawa, S.; Murao, S.; Sakurai, S.

    2004-01-01

    We recently developed and reported a three-detector measuring system making use of a pure-Ge detector combined with two Si(Li) detectors. The efficiency curve of the pure-Ge detector was determined as relative efficiencies to those of the existing Si(Li) detectors and accuracy of it was confirmed by analyzing a few samples whose elemental concentrations were known. It was found that detection of fluorine becomes possible by analyzing prompt γ-rays and the detection limit was found to be less than 0.1 ppm for water samples. In this work, a method of quantitative analysis of fluorine has been established in order to investigate environmental contamination by fluorine. This method is based on the fact that both characteristic x-rays from many elements and 110 keV prompt γ-rays from fluorine can be detected in the same spectrum. The present method is applied to analyses of a few environmental samples such as tealeaves, feed for domestic animals and human bone. The results are consistent with those obtained by other methods and it is found that the present method is quite useful and convenient for investigation studies on regional pollution by fluorine. (author)

  11. The scope of detector Medipix2 in micro-radiography of biological samples

    International Nuclear Information System (INIS)

    Dammer, J.; Weyda, F.; Jakubek, J.; Skrabal, P.; Sopko, V.; Vavrik, D.

    2011-01-01

    We present our experimental setup devoted to high resolution X-ray micro-radiography that is suitable for imaging of small biological samples. The photon source is a FeinFocus micro-focus X-ray tube. The single photon counting pixel device Medipix2 serves as imaging area. Recently used imaging detectors as radiography films or scintillator detectors, cannot visualize required information about inner structure of scanned sample. Detectors Medipix2 do not suffer from so-called dark current noise and work in unlimited dynamic range. These features of detectors confer high quality and high contrast of final images. The radiographic imaging with detectors Medipix2 represents non-invasive and non-destructive method of investigation. Hereby, we demonstrate results of micro-radiographic study of internal structures of tiny biological samples. In addition to morphological and anatomical studies, we would like to present preliminary study of dynamic processes inside of organisms using micro-radiographic video-capturing.

  12. The scope of detector Medipix2 in micro-radiography of biological samples

    Energy Technology Data Exchange (ETDEWEB)

    Dammer, J., E-mail: jiri.dammer@utef.cvut.cz [Institute of Experimental and Applied Physics, Czech Technical University in Prague, Horska 3a/22, CZ-12800 Prague 2 (Czech Republic); Weyda, F. [Biology Centre of the Academy of Sciences of the Czech Republic, Institute of Entomology, Branisovska 31, CZ-37005 Ceske Budejovice (Czech Republic); Faculty of Science, University of South Bohemia, Branisovska 31, CZ-37005 Ceske Budejovice (Czech Republic); Jakubek, J. [Institute of Experimental and Applied Physics, Czech Technical University in Prague, Horska 3a/22, CZ-12800 Prague 2 (Czech Republic); Skrabal, P. [Faculty of Biomedical Engineering, Czech Technical University in Prague, Nam. Sitna 3105, CZ-272 01 Kladno (Czech Republic); Sopko, V.; Vavrik, D. [Institute of Experimental and Applied Physics, Czech Technical University in Prague, Horska 3a/22, CZ-12800 Prague 2 (Czech Republic)

    2011-05-15

    We present our experimental setup devoted to high resolution X-ray micro-radiography that is suitable for imaging of small biological samples. The photon source is a FeinFocus micro-focus X-ray tube. The single photon counting pixel device Medipix2 serves as imaging area. Recently used imaging detectors as radiography films or scintillator detectors, cannot visualize required information about inner structure of scanned sample. Detectors Medipix2 do not suffer from so-called dark current noise and work in unlimited dynamic range. These features of detectors confer high quality and high contrast of final images. The radiographic imaging with detectors Medipix2 represents non-invasive and non-destructive method of investigation. Hereby, we demonstrate results of micro-radiographic study of internal structures of tiny biological samples. In addition to morphological and anatomical studies, we would like to present preliminary study of dynamic processes inside of organisms using micro-radiographic video-capturing.

  13. Manual for the Epithermal Neutron Multiplicity Detector (ENMC) for Measurement of Impure MOX and Plutonium Samples

    International Nuclear Information System (INIS)

    Menlove, H. O.; Rael, C. D.; Kroncke, K. E.; DeAguero, K. J.

    2004-01-01

    We have designed a high-efficiency neutron detector for passive neutron coincidence and multiplicity counting of dirty scrap and bulk samples of plutonium. The counter will be used for the measurement of impure plutonium samples at the JNC MOX fabrication facility in Japan. The counter can also be used to create working standards from bulk process MOX. The detector uses advanced design "3He tubes to increase the efficiency and to shorten the neutron die-away time. The efficiency is 64% and the die-away time is 19.1 ?s. The Epithermal Neutron Multiplicity Counter (ENMC) is designed for high-precision measurements of bulk plutonium samples with diameters of less than 200 mm. The average neutron energy from the sample can be measured using the ratio of the inner ring of He-3 tubes to the outer ring. This report describes the hardware, performance, and calibration for the ENMC.

  14. Application of digital sampling techniques to particle identification in scintillation detectors

    International Nuclear Information System (INIS)

    Bardelli, L.; Bini, M.; Poggi, G.; Taccetti, N.

    2002-01-01

    In this paper, the use of a fast digitizing system for identification of fast charged particles with scintillation detectors is discussed. The three-layer phoswich detectors developed in the framework of the FIASCO experiment for the detection of light charged particles (LCP) and intermediate mass fragments (IMF) emitted in heavy-ion collisions at Fermi energies are briefly discussed. The standard analog electronics treatment of the signals for particle identification is illustrated. After a description of the digitizer designed to perform a fast digital sampling of the phoswich signals, the feasibility of particle identification on the sampled data is demonstrated. The results obtained with two different pulse shape discrimination analyses based on the digitally sampled data are compared with the standard analog signal treatment. The obtained results suggest, for the present application, the replacement of the analog methods with the digital sampling technique

  15. Self-absorption corrections of various sample-detector geometries in gamma-ray spectrometry using sample Monte Carlo Simulations

    International Nuclear Information System (INIS)

    Ahmad Saat; Appleby, P.G.; Nolan, P.J.

    1997-01-01

    Corrections for self-absorption in gamma-ray spectrometry have been developed using a simple Monte Carlo simulation technique. The simulation enables the calculation of gamma-ray path lengths in the sample which, using available data, can be used to calculate self-absorption correction factors. The simulation was carried out on three sample geometries: disk, Marinelli beaker, and cylinder (for well-type detectors). Mathematical models and experimental measurements are used to evaluate the simulations. A good agreement of within a few percents was observed. The simulation results are also in good agreement with those reported in the literature. The simulation code was carried out in FORTRAN 90,

  16. Contributions for the application of a phoswich detector on the analysis of environmental samples

    International Nuclear Information System (INIS)

    Dalaqua Junior, L.

    1989-01-01

    The characteristics of a phoswich detector and the parameters of the pulse shape descrimination system are evaluated aiming the application on environmental analysis by direct low level gamma ray spectrometry. The calibration curves and adjustments for the pulse discrimination, detector resolution and homogeneity measurements are presented. Background reduction and the 210 Pb detection eficiency on evaporated sources are evaluated. The results obtained demonstrates the application potentiality on the analysis of environmental samples due to a high detection eficiency and good geometry conditions to the measurements. (author) [pt

  17. Study on the etching conditions of polycarbonate detectors for particle analysis of safeguards environmental samples

    International Nuclear Information System (INIS)

    Iguchi, K.; Esaka, K.T.; Lee, C.G.; Inagawa, J.; Esaka, F.; Onodera, T.; Fukuyama, H.; Suzuki, D.; Sakurai, S.; Watanabe, K.; Usuda, S.

    2005-01-01

    The fission track technique was applied to the particle analysis for safeguards environmental samples to obtain information about the isotope ratio of nuclear materials in individual particles. To detect the particles containing nuclear material with high detection efficiency and less particle loss, the influence of uranium enrichments on etching conditions of a fission track detector made of polycarbonate was investigated. It was shown that the increase in uranium enrichment shortened the suitable etching time both for particle detection and for less particle loss. From the results obtained, it was suggested that the screening of the uranium particles according to the enrichment is possible by controlling the etching time of the detector

  18. Shielding practice

    International Nuclear Information System (INIS)

    Sauermann, P.F.

    1985-08-01

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

  19. Calibration of track detectors and measurement of radon exhalation rate from solid samples

    International Nuclear Information System (INIS)

    Singh, Ajay Kumar; Jojo, P.J.; Prasad, Rajendra; Khan, A.J.; Ramachandran, T.V.

    1997-01-01

    CR-39 and LR-115 type II track detectors to be used for radon exhalation measurements have been calibrated. The configurations fitted with detectors in Can technique in the open cup mode are cylindrical plastic cup (PC) and conical plastic cup (CPC). The experiment was performed in radon exposure chamber having monodisperse aerosols of 0.2 μm size, to find the relationship between track density and the radon concentration. The calibration factors for PC and CPC type dosimeters with LR-115 type II detector were found to be 0.056 and 0.083 tracks cm -2 d -1 (Bqm -3 ) -1 respectively, while with CR-39 detector the values were 0.149 and 0.150 tracks cm -2 d -1 (Bq m -3 ) -1 . Employing the Can technique, measurements of exhalation rates from solid samples used as construction materials, are undertaken. Radon exhalation rate is found to be minimum in cement samples while in fly ash it is not enhanced as compared to coal samples. (author)

  20. New shielding material development for compact accelerator-driven neutron source

    Directory of Open Access Journals (Sweden)

    Guang Hu

    2017-04-01

    Full Text Available The Compact Accelerator-driven Neutron Source (CANS, especially the transportable neutron source is longing for high effectiveness shielding material. For this reason, new shielding material is researched in this investigation. The component of shielding material is designed and many samples are manufactured. Then the attenuation detection experiments were carried out. In the detections, the dead time of the detector appeases when the proton beam is too strong. To grasp the linear range and nonlinear range of the detector, two currents of proton are employed in Pb attenuation detections. The transmission ratio of new shielding material, polyethylene (PE, PE + Pb, BPE + Pb is detected under suitable current of proton. Since the results of experimental neutrons and γ-rays appear as together, the MCNP and PHITS simulations are applied to assisting the analysis. The new shielding material could reduce of the weight and volume compared with BPE + Pb and PE + Pb.

  1. Optimization of the two-sample rank Neyman-Pearson detector

    Science.gov (United States)

    Akimov, P. S.; Barashkov, V. M.

    1984-10-01

    The development of optimal algorithms concerned with rank considerations in the case of finite sample sizes involves considerable mathematical difficulties. The present investigation provides results related to the design and the analysis of an optimal rank detector based on a utilization of the Neyman-Pearson criteria. The detection of a signal in the presence of background noise is considered, taking into account n observations (readings) x1, x2, ... xn in the experimental communications channel. The computation of the value of the rank of an observation is calculated on the basis of relations between x and the variable y, representing interference. Attention is given to conditions in the absence of a signal, the probability of the detection of an arriving signal, details regarding the utilization of the Neyman-Pearson criteria, the scheme of an optimal rank, multichannel, incoherent detector, and an analysis of the detector.

  2. Shielding research in France

    Energy Technology Data Exchange (ETDEWEB)

    Lafore, P

    1964-10-01

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

  3. Shielding in experimental areas

    International Nuclear Information System (INIS)

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  4. Electromagnetic shielding

    International Nuclear Information System (INIS)

    Tzeng, Wen-Shian V.

    1991-01-01

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

  5. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  6. High-energy detector

    Science.gov (United States)

    Bolotnikov, Aleksey E [South Setauket, NY; Camarda, Giuseppe [Farmingville, NY; Cui, Yonggang [Upton, NY; James, Ralph B [Ridge, NY

    2011-11-22

    The preferred embodiments are directed to a high-energy detector that is electrically shielded using an anode, a cathode, and a conducting shield to substantially reduce or eliminate electrically unshielded area. The anode and the cathode are disposed at opposite ends of the detector and the conducting shield substantially surrounds at least a portion of the longitudinal surface of the detector. The conducting shield extends longitudinally to the anode end of the detector and substantially surrounds at least a portion of the detector. Signals read from one or more of the anode, cathode, and conducting shield can be used to determine the number of electrons that are liberated as a result of high-energy particles impinge on the detector. A correction technique can be implemented to correct for liberated electron that become trapped to improve the energy resolution of the high-energy detectors disclosed herein.

  7. Automatic setting of the distance between sample and detector in gamma-ray spectroscopy

    International Nuclear Information System (INIS)

    Andeweg, A.H.

    1980-01-01

    An apparatus has been developed that automatically sets the distance from the sample to the detector according to the radioactivity of the sample. The distance-setting unit works in conjuction with an automatic sample changer, and is interconnected with other components so that the counting head automatically moves to the optimum distance for the analysis of a particular sample. The distance, which is indicated digitally in increments of 0,01 mm, can be set between 18 and 995 mm at count rates that can be preset between 1000 and 10 000 counts per second. On being tested, the instrument performed well within the desired range and accuracy. Under routine conditions, the spectra were much more accurate than before, especially when samples of different radioactivity were counted

  8. Matching Ge detector element geometry to sample size and shape: One does not fit all exclamation point

    International Nuclear Information System (INIS)

    Keyser, R.M.; Twomey, T.R.; Sangsingkeow, P.

    1998-01-01

    For 25 yr, coaxial germanium detector performance has been specified using the methods and values specified in Ref. 1. These specifications are the full-width at half-maximum (FWHM), FW.1M, FW.02M, peak-to-Compton ratio, and relative efficiency. All of these measurements are made with a 60 Co source 25 cm from the cryostat endcap and centered on the axis of the detector. These measurements are easy to reproduce, both because they are simple to set up and use a common source. These standard tests have been useful in guiding the user to an appropriate detector choice for the intended measurement. Most users of germanium gamma-ray detectors do not make measurements in this simple geometry. Germanium detector manufacturers have worked over the years to make detectors with better resolution, better peak-to-Compton ratios, and higher efficiency--but all based on measurements using the IEEE standard. Advances in germanium crystal growth techniques have made it relatively easy to provide detector elements of different shapes and sizes. Many of these different shapes and sizes can give better results for a specific application than other shapes and sizes. But, the detector specifications must be changed to correspond to the actual application. Both the expected values and the actual parameters to be specified should be changed. In many cases, detection efficiency, peak shape, and minimum detectable limit for a particular detector/sample combination are valuable specifications of detector performance. For other situations, other parameters are important, such as peak shape as a function of count rate. In this work, different sample geometries were considered. The results show the variation in efficiency with energy for all of these sample and detector geometries. The point source at 25 cm from the endcap measurement allows the results to be compared with the currently given IEEE criteria. The best sample/detector configuration for a specific measurement requires more and

  9. A new TXRF vacuum chamber with sample changer for chemical analysis using silicon drift chamber detector

    International Nuclear Information System (INIS)

    Streli, C.; Wobrauschek, P.; Zoeger, N.; Pepponi, G.

    2003-01-01

    Full text: Several TXRF spectrometers for chemical analysis as well as for wafer surface analysis are commercially available. But there is no one available for chemical analysis offering the possibility to measure the samples in vacuum conditions. Vacuum of 10 -2 mbar in the sample environment helps to reduce the background due to scattering from air, thus to improve the detection limits as well as to reduce the absorption of low energy fluorescence radiation from low Z elements and extend the elemental range to be measured and removes the Ar lines from the spectrum. The x-ray group of the Atominstitut designed and fabricated a new vacuum chamber for TXRF equipped with a 12 position sample changer from Italstructures, Riva, Italy. The detector used was a 10 mm 2 silicon drift detector (KETEK, Munich, Germany), offering the advantage of electrically cooling, so no LN2 is required. The chamber was designed to be attached to a diffraction tube housing, e.g. with a fine focus Mo-x-ray tube and uses a multilayer monochromator. Spectra are stored by a small AMTEK MCA and control between sample changer and MCA communication is done by a modified AMPTEK software. The performance is expressed in detection limits of 1 pg Rb for Mo Ka excitation with 50 kV, 40 mA excitation conditions, 1000 s lifetime, obtained from a sample containing 600 pg Rb as single element standard. Details on performance, reproducibility and light element excitation and detection are presented. (author)

  10. Shielding plugs

    International Nuclear Information System (INIS)

    Makishima, Kenji.

    1986-01-01

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

  11. Sampling theorem for geometric moment determination and its application to a laser beam position detector.

    Science.gov (United States)

    Loce, R P; Jodoin, R E

    1990-09-10

    Using the tools of Fourier analysis, a sampling requirement is derived that assures that sufficient information is contained within the samples of a distribution to calculate accurately geometric moments of that distribution. The derivation follows the standard textbook derivation of the Whittaker-Shannon sampling theorem, which is used for reconstruction, but further insight leads to a coarser minimum sampling interval for moment determination. The need for fewer samples to determine moments agrees with intuition since less information should be required to determine a characteristic of a distribution compared with that required to construct the distribution. A formula for calculation of the moments from these samples is also derived. A numerical analysis is performed to quantify the accuracy of the calculated first moment for practical nonideal sampling conditions. The theory is applied to a high speed laser beam position detector, which uses the normalized first moment to measure raster line positional accuracy in a laser printer. The effects of the laser irradiance profile, sampling aperture, number of samples acquired, quantization, and noise are taken into account.

  12. Study of Efficiency Calibrations of HPGe Detectors for Radioactivity Measurements of Environmental Samples

    International Nuclear Information System (INIS)

    Harb, S.; Salahel Din, K.; Abbady, A.

    2009-01-01

    In this paper, we describe a method of calibrating of efficiency of a HPGe gamma-ray spectrometry of bulk environmental samples (Tea, crops, water, and soil) is a significant part of the environmental radioactivity measurements. Here we will discuss the full energy peak efficiency (FEPE) of three HPGe detectors it as a consequence, it is essential that the efficiency is determined for each set-up employed. Besides to take full advantage at gamma-ray spectrometry, a set of efficiency at several energies which covers the wide the range in energy, the large the number of radionuclides whose concentration can be determined to measure the main natural gamma-ray emitters, the efficiency should be known at least from 46.54 keV ( 210 Pb) to 1836 keV ( 88 Y). Radioactive sources were prepared from two different standards, a first mixed standard QC Y 40 containing 210 Pb, 241 Am, 109 Cd, and Co 57 , and the second QC Y 48 containing 241 Am, 109 Cd, 57 Co, 139 Ce, 113 Sn, 85 Sr, 137 Cs, 88 Y, and 60 Co is necessary in order to calculate the activity of the different radionuclides contained in a sample. In this work, we will study the efficiency calibration as a function of different parameters as:- Energy of gamma ray from 46.54 keV ( 210 Pb) to 1836 keV ( 88 Y), three different detectors A, B, and C, geometry of containers (point source, marinelli beaker, and cylindrical bottle 1 L), height of standard soil samples in bottle 250 ml, and density of standard environmental samples. These standard environmental sample must be measured before added standard solution because we will use the same environmental samples in order to consider the self absorption especially and composition in the case of volume samples.

  13. Detection of fission fragments using thick samples in contact with solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Lima, D.A. de; Martins, J.B.; Tavares, O.A.P.

    1987-01-01

    Whenever use is made of thick samples in contact with solid state nuclear track detectors for determining fission yields, one of the fundamental problems is the evaluation of the effective number of target nuclei which contributes to the fraction of the number of fission events that will be recorded. The evaluation of the effective number of target nuclei which contributes to recorded events is based on the effective thickness of the sample. A method for evaluating effective thickness of thick samples for binary fission modes, is presented. A cross section equation which takes into account all the necessary corrections due to fragment attenuation effects by a thick target for calculation induced fission yields, was obtained. (Author) [pt

  14. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0νββ experiment

    International Nuclear Information System (INIS)

    Peiffer, J.P.

    2007-01-01

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0νββ) decay, which can be investigated using 76 Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0νββ decay in the Heidelberg-Moscow (HdM) 76 Ge experiment by a part of the HdM collaboration. The new 76 Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of ≤10 -2 events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of ≤10 -3 events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in 76 Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced 60 Co is expected to be ∝2.5.10 -3 events/(kg.keV.y). LAr scintillates in UV (λ=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for 60 Co and a factor 17 for 232 Th around Q ββ = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for 232 Th and 60 Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength shifter/reflector combination for the light detection has been developed, leading to a photo electron (pe) yield of as much as

  15. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0{nu}{beta}{beta} experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peiffer, J.P.

    2007-07-25

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0{nu}{beta}{beta}) decay, which can be investigated using {sup 76}Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0{nu}{beta}{beta} decay in the Heidelberg-Moscow (HdM) {sup 76}Ge experiment by a part of the HdM collaboration. The new {sup 76}Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of {<=}10{sup -2} events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of {<=}10{sup -3} events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in {sup 76}Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced {sup 60}Co is expected to be {proportional_to}2.5.10{sup -3} events/(kg.keV.y). LAr scintillates in UV ({lambda}=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for {sup 60}Co and a factor 17 for {sup 232}Th around Q{sub {beta}}{sub {beta}} = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for {sup 232}Th and {sup 60}Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength

  16. Dosimetry and shielding

    International Nuclear Information System (INIS)

    Farinelli, U.

    1977-01-01

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

  17. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  18. Sample dependent response of a LaCl{sub 3}:Ce detector in prompt gamma neutron activation analysis of bulk hydrocarbon samples

    Energy Technology Data Exchange (ETDEWEB)

    Naqvi, A.A., E-mail: aanaqvi@kfupm.edu.sa [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Al-Matouq, Faris A.; Khiari, F.Z. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Isab, A.A. [Department of Chemistry, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Khateeb-ur-Rehman,; Raashid, M. [Department of Physics, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia)

    2013-08-11

    The response of a LaCl{sub 3}:Ce detector has been found to depend upon the hydrogen content of bulk samples in prompt gamma analysis using 14 MeV neutron inelastic scattering. The moderation of 14 MeV neutrons from hydrogen in the bulk sample produces thermal neutrons around the sample which ultimately excite chlorine capture gamma rays in the LaCl{sub 3}:Ce detector material. Interference of 6.11 MeV chlorine gamma rays from the detector itself with 6.13 MeV oxygen gamma rays from the bulk samples makes the intensity of the 6.13 MeV oxygen gamma ray peak relatively insensitive to variations in oxygen concentration. The strong dependence of the 1.95 MeV doublet chlorine gamma ray yield on hydrogen content of the bulk samples confirms fast neutron moderation from hydrogen in the bulk samples as a major source of production of thermal neutrons and chlorine gamma rays in the LaCl{sub 3}:Ce detector material. Despite their poor oxygen detection capabilities, these detectors have nonetheless excellent detection capabilities for hydrogen and carbon in benzene, butyl alcohol, propanol, propanic acid, and formic acid bulk samples using 14 MeV neutron inelastic scattering.

  19. Experimental technique to measure thoron generation rate of building material samples using RAD7 detector

    International Nuclear Information System (INIS)

    Csige, I.; Szabó, Zs.; Szabó, Cs.

    2013-01-01

    Thoron ( 220 Rn) is the second most abundant radon isotope in our living environment. In some dwellings it is present in significant amount which calls for its identification and remediation. Indoor thoron originates mainly from building materials. In this work we have developed and tested an experimental technique to measure thoron generation rate in building material samples using RAD7 radon-thoron detector. The mathematical model of the measurement technique provides the thoron concentration response of RAD7 as a function of the sample thickness. For experimental validation of the technique an adobe building material sample was selected for measuring the thoron concentration at nineteen different sample thicknesses. Fitting the parameters of the model to the measurement results, both the generation rate and the diffusion length of thoron was estimated. We have also determined the optimal sample thickness for estimating the thoron generation rate from a single measurement. -- Highlights: • RAD7 is used for the determination of thoron generation rate (emanation). • The described model takes into account the thoron decay and attenuation. • The model describes well the experimental results. • A single point measurement method is offered at a determined sample thickness

  20. Gonadal Shielding in Radiography: A Best Practice?

    Science.gov (United States)

    Fauber, Terri L

    2016-11-01

    To investigate radiation dose to phantom testes with and without shielding. A male anthropomorphic pelvis phantom was imaged with thermoluminescent dosimeters (TLDs) placed in the right and left detector holes corresponding to the testes. Ten exposures were made of the pelvis with and without shielding. The exposed TLDs were packaged securely and mailed to the University of Wisconsin Calibration Laboratory for reading and analysis. A t test was calculated for the 2 exposure groups (no shield and shielded) and found to be significant, F = 8.306, P shield was used during pelvic imaging. Using a flat contact shield during imaging of the adult male pelvis significantly reduces radiation dose to the testes. Regardless of the contradictions in the literature on gonadal shielding, the routine practice of shielding adult male gonads during radiographic imaging of the pelvis is a best practice. © 2016 American Society of Radiologic Technologists.

  1. Electromagnetic shield

    International Nuclear Information System (INIS)

    Miller, J.S.

    1987-01-01

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

  2. Neutron shieldings

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1979-01-01

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

  3. Nuclear shields

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  4. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  5. Uranium concentration in blood samples of Southern Iraqi leukemia patients using CR-39 track detector

    International Nuclear Information System (INIS)

    Al-Hamzawi, A.A.; Al-Qadisiyah University, Qadisiyah; Jaafar, M.S.; Tawfiq, N.F.

    2014-01-01

    The simple and effective technique of fission track etch has been applied to determine trace concentration of uranium in human blood samples taken from two groups of male and female participants: leukemia patients and healthy subjects group. The blood samples of leukemia patients and healthy subjects were collected from three key southern governorates namely, Basrah, Muthanna and Dhi-Qar. These governorates were the centers of intensive military activities during the 1991 and 2003 Gulf wars, and the discarded weapons are still lying around in these regions. CR-39 track detector was used for registration of induced fission tracks. The results show that the highest recorded uranium concentration in the blood samples of leukemia patients was 4.71 ppb (female, 45 years old, from Basrah) and the minimum concentration was 1.91 ppb (male, 3 years old, from Muthanna). For healthy group, the maximum uranium concentration was 2.15 ppb (female, 55 years old, from Basrah) and the minimum concentration was 0.86 ppb (male, 5 years old, from Dhi-Qar). It has been found that the uranium concentrations in human blood samples of leukemia patients are higher than those of the healthy group. These uranium concentrations in the leukemia patients group were significantly different (P < 0.001) from those in the healthy group. (author)

  6. A comparative study of shadow shield whole body monitor incorporated with 203 mm dia. x 102 mm thick and 102 mm dia. x 76 mm thick NaI(TI) detectors

    International Nuclear Information System (INIS)

    Sankhla, Rajesh; Singh, I.S.; Rao, D.D.

    2016-01-01

    The whole body counting using Shadow Shield Whole Body Monitor (SSWBM) proved to be a popular method for assessment of internal contamination due to high energy gamma (E>200 keV) emitting radio nuclides that got inadvertently incorporated in the occupational workers. Currently ∼ 5 SSWBMs are operational at various DAE nuclear facilities throughout the country. The shielding of SSWBMs are said to be designed for 102 mm x 76 mm NaI(Tl) detector and over a period of time, the same concept is being followed. At present, the number of subjects monitored per annum has increased significantly compared to earlier years due to the increase in nuclear facilities at different sites and also increase in number of contract personnel. Aim of this study is to develop/upgrade the existing SSWBMs to increase their capabilities in terms of throughput without compromising on sensitivity. This work includes response studies of individual detectors of sizes 102 mm x 76 mm and 203 mm x 102 mm housed in SSWBM in terms of background, efficiency and Minimum Detection Activity (MDA) for different gamma emitting radio nuclides using Bhabha Atomic Research Centre reference Bottle Mannequin ABsorption (BOMAB) phantom

  7. Calculation of depleted uranium concentration in dental fillings samples using the nuclear track detector CR-39

    International Nuclear Information System (INIS)

    Mahdi, K. H.; Subhi, A. T.; Tawfiq, N. F.

    2012-12-01

    The purpose of this study is to determine the concentration of depleted uranium in dental fillings samples, which were obtained some hospital and dental office, sale of materials deployed in Iraq. 8 samples were examined from two different fillings and lead-filling (amalgam) and composite filling (plastic). concentrations of depleted uranium were determined in these samples using a nuclear track detector CR-39 through the recording of the tracks left by of fragments of fission resulting from the reaction 2 38U (n, f). The samples are bombarded by neutrons emitted from the neutron source (2 41A m-Be) with flux of ( 10 5 n. cm- 2. s -1 ). The period of etching to show the track of fission fragments is 5 hours using NaOH solution with normalization (6.25N), and temperature (60 o C ). Concentration of depleted uranium were calculated by comparison with standard samples. The result that obtained showed that the value of the weighted average for concentration of uranium in the samples fillings (5.54± 1.05) ppm lead to thr filling (amalgam) and (5.33±0.6) ppm of the filling composite (plastic). The hazard- index, the absorbed dose and the effective dose for these concentration were determined. The obtained results of the effective dose for each of the surface of the bone and skin (as the areas most affected by this compensation industrial) is (0.56 mSv / y) for the batting lead (amalgam) and (0.54 mSv / y) for the filling composite (plastic). From the results of study it was that the highest rate is the effective dose to a specimen amalgam filling (0.68 mSv / y) which is less than the allowable limit for exposure of the general people set the World Health Organization (WHO), a (1 mSv / y). (Author)

  8. Si(Li)-NaI(Tl) sandwich detector array for measurements of trace radionuclides in soil samples

    International Nuclear Information System (INIS)

    Strauss, M.G.; Sherman, I.S.; Roche, C.T.; Pehl, R.H.

    1986-01-01

    An ultra-sensitive X/γ-ray detector system for assaying trace radioactivity in actinide contaminated soil and ash samples has been developed. The new system consists of an array of 6 large Si(Li) X-ray detectors sensitive on both faces and mounted on edge in a paddle-shaped cryostat with a 14 cm diameter Be window on each side. The paddle, with a sample of the soil placed at each window, is sandwiched between 2 large NaI(Tl) scintillators which suppress the γ background. With X-rays being measured simultaneously from soil in 2 sample holders and background reduced by 50% using anticoincidence, the sensitivity of this detector is 4 times higher than that of conventionally mounted Si(Li) detectors. A soil sample containing 50 pCi/g 239 Pu was measured in 5 min with an uncertainty of 1 and NpLsub(β1) X-ray peaks are resolved thus permitting measurement of trace Pu in the presence of 241 Am. This is the most sensitive and selective detector known for nondestructive assay of radioactivity in soil and other samples. (orig.)

  9. Neutron activation measurements in research reactor concrete shield

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Bozic, M.

    2001-01-01

    The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

  10. Detector Sampling of Optical/IR Spectra: How Many Pixels per FWHM?

    Science.gov (United States)

    Robertson, J. Gordon

    2017-08-01

    Most optical and IR spectra are now acquired using detectors with finite-width pixels in a square array. Each pixel records the received intensity integrated over its own area, and pixels are separated by the array pitch. This paper examines the effects of such pixellation, using computed simulations to illustrate the effects which most concern the astronomer end-user. It is shown that coarse sampling increases the random noise errors in wavelength by typically 10-20 % at 2 pixels per Full Width at Half Maximum, but with wide variation depending on the functional form of the instrumental Line Spread Function (i.e. the instrumental response to a monochromatic input) and on the pixel phase. If line widths are determined, they are even more strongly affected at low sampling frequencies. However, the noise in fitted peak amplitudes is minimally affected by pixellation, with increases less than about 5%. Pixellation has a substantial but complex effect on the ability to see a relative minimum between two closely spaced peaks (or relative maximum between two absorption lines). The consistent scale of resolving power presented by Robertson to overcome the inadequacy of the Full Width at Half Maximum as a resolution measure is here extended to cover pixellated spectra. The systematic bias errors in wavelength introduced by pixellation, independent of signal/noise ratio, are examined. While they may be negligible for smooth well-sampled symmetric Line Spread Functions, they are very sensitive to asymmetry and high spatial frequency sub-structure. The Modulation Transfer Function for sampled data is shown to give a useful indication of the extent of improperly sampled signal in an Line Spread Function. The common maxim that 2 pixels per Full Width at Half Maximum is the Nyquist limit is incorrect and most Line Spread Functions will exhibit some aliasing at this sample frequency. While 2 pixels per Full Width at Half Maximum is nevertheless often an acceptable minimum for

  11. MicroShield/ISOCS gamma modeling comparison.

    Energy Technology Data Exchange (ETDEWEB)

    Sansone, Kenneth R

    2013-08-01

    Quantitative radiological analysis attempts to determine the quantity of activity or concentration of specific radionuclide(s) in a sample. Based upon the certified standards that are used to calibrate gamma spectral detectors, geometric similarities between sample shape and the calibration standards determine if the analysis results developed are qualitative or quantitative. A sample analyzed that does not mimic a calibrated sample geometry must be reported as a non-standard geometry and thus the results are considered qualitative and not quantitative. MicroShieldR or ISOCSR calibration software can be used to model non-standard geometric sample shapes in an effort to obtain a quantitative analytical result. MicroShieldR and Canberras ISOCSR software contain several geometry templates that can provide accurate quantitative modeling for a variety of sample configurations. Included in the software are computational algorithms that are used to develop and calculate energy efficiency values for the modeled sample geometry which can then be used with conventional analysis methodology to calculate the result. The response of the analytical method and the sensitivity of the mechanical and electronic equipment to the radionuclide of interest must be calibrated, or standardized, using a calibrated radiological source that contains a known and certified amount of activity.

  12. Measurement of radon-222 concentration in environment sampled within short time using charcoal detector

    International Nuclear Information System (INIS)

    Yamasaki, Tadashi; Sekiyama, Shigenobu; Tokin, Mina; Nakayasu, Yumiko; Watanabe, Tamaki.

    1994-01-01

    The concentration of 222 Rn in air sampled within a very short period of time was measured using activated charcoal as the adsorber. The detector is the plastic canister containing mixture of the activated charcoal and the silica gel. The radon gas was adsorbed in the charcoal in the radon chamber at the temperature of 25degC. A little amount of liquid scintillation cocktail was added into the vial of liquid scintillation counter with the canister. The radon in the charcoal was extracted in the liquid scintillation cocktail. Alpha particles emitted from radon and its daughter nuclei in the cocktail were detected using the liquid scintillation counter. Present method has advantages of not only short sampling time of air but also adsorption of radon in charcoal under a constant temperature. The concentration of radon in air down to 2 Bq/m 3 could be detected. A kinetic model for adsorption of radon in the charcoal is also presented. The ratio of radon concentration in the charcoal to that in air under the equilibrium state of adsorption was estimated to be from 6.1 to 6.8 m 3 /kg at the temperature of 25degC. (author)

  13. Radiation shielding

    International Nuclear Information System (INIS)

    Aitken, D.

    1979-01-01

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

  14. A framework for inference about carnivore density from unstructured spatial sampling of scat using detector dogs

    Science.gov (United States)

    Thompson, Craig M.; Royle, J. Andrew; Garner, James D.

    2012-01-01

    Wildlife management often hinges upon an accurate assessment of population density. Although undeniably useful, many of the traditional approaches to density estimation such as visual counts, livetrapping, or mark–recapture suffer from a suite of methodological and analytical weaknesses. Rare, secretive, or highly mobile species exacerbate these problems through the reality of small sample sizes and movement on and off study sites. In response to these difficulties, there is growing interest in the use of non-invasive survey techniques, which provide the opportunity to collect larger samples with minimal increases in effort, as well as the application of analytical frameworks that are not reliant on large sample size arguments. One promising survey technique, the use of scat detecting dogs, offers a greatly enhanced probability of detection while at the same time generating new difficulties with respect to non-standard survey routes, variable search intensity, and the lack of a fixed survey point for characterizing non-detection. In order to account for these issues, we modified an existing spatially explicit, capture–recapture model for camera trap data to account for variable search intensity and the lack of fixed, georeferenced trap locations. We applied this modified model to a fisher (Martes pennanti) dataset from the Sierra National Forest, California, and compared the results (12.3 fishers/100 km2) to more traditional density estimates. We then evaluated model performance using simulations at 3 levels of population density. Simulation results indicated that estimates based on the posterior mode were relatively unbiased. We believe that this approach provides a flexible analytical framework for reconciling the inconsistencies between detector dog survey data and density estimation procedures.

  15. Radiation shielding quality assurance

    Science.gov (United States)

    Um, Dallsun

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

  16. Gonadal shield.

    Science.gov (United States)

    Purdy, J A; Stiteler, R D; Glasgow, G P; Mill, W B

    1975-10-01

    A secondary gonadal shield for use in the pelvic irradiation of males was designed and built using material and apparatus available with the Cerrobend blocking system. The gonadal dose was reduced to approximately 1.5 to 2.5% of the given dose.

  17. Hydrogen, carbon and oxygen determination in proxy material samples using a LaBr3:Ce detector

    International Nuclear Information System (INIS)

    Naqvi, A.A.; Al-Matouq, Faris A.; Khiari, F.Z.; Isab, A.A.; Raashid, M.; Khateeb-ur-Rehman

    2013-01-01

    Hydrogen, carbon and oxygen concentrations were measured in caffeine, urea, ammonium acetate and melamine bulk samples via 14 MeV neutron inelastic scattering using a LaBr 3 :Ce detector. The samples tested herein represent drugs, explosives and benign materials, respectively. Despite its intrinsic activity, the LaBr 3 :Ce detector performed well in detecting the hydrogen, carbon and oxygen elements. Because 5.1 MeV nitrogen gamma rays interfere with silicon and calcium prompt gamma rays from the room background, the nitrogen peak was not detected in the samples. An excellent agreement was observed between the experimental and theoretical yields of 2.22, 4.43 and 6.13 MeV gamma rays from the analyzed samples as a function of H, C and O concentrations, respectively. Within statistical errors, the minimum detectable concentration (MDC) of hydrogen, carbon and oxygen elements in the tested materials were consistent with previously reported MDC values for these elements measured in hydrocarbon samples. - Highlights: • Hydrogen, carbon and oxygen concentration measurement in bulk samples using 14 MeV neutrons induced prompt gamma rays. • Prompt gamma analysis of narcotics and explosive proxy materials e.g. ammonium acetate, caffeine, urea and melamine Bulk samples. • Prompt gamma detection using large cylindrical 76×76 mm 2 (diameter x height ) LaBr 3 :Ce detector. • Carbon/oxygen elemental ratio measurement from explosive and narcotics proxy material samples

  18. Physical properties and petrologic description of rock samples from an IOCG mineralized area in the northern Fennoscandian Shield, Sweden

    DEFF Research Database (Denmark)

    Sandrin, Alessandro; Edfelt, Å.; Waight, Tod Earle

    2009-01-01

    The Tjårrojåkka Fe-Cu prospect in northern Sweden is considered an example of a Fe-oxide Cu-Au (IOCG) deposit and is hosted in metamorphosed Paleoproterozoic volcanic and intrusive rocks. Rock samples from 24 outcrops were collected for petrophysical analysis (magnetic susceptibility, remanent ma...

  19. Investigation of (n, 2n) reaction and fission rates in iron-shielded uranium samples bombarded by 14.9 MeV neutrons

    International Nuclear Information System (INIS)

    Shani, G.

    1976-01-01

    The effect of the thickness of iron shielding on the (n, 2n) reaction rate in a fusion reactor (hybrid) blanket is investigated. The results are compared with the fission rate-dependence. Samples of natural uranium are irradiated with 14 MeV neutrons, with iron slabs of various thickness between the neutron generator target and the samples. Both reactions are threshold reactions but the fact that the 238 U (n, 2n) reaction threshold is at 6 MeV and that of fission is at 2 MeV makes the ratio between the two very much geometry-dependent. Two geometrical effects take place, the 1/r 2 and the build-up. While the build-up affects the (n, 2n) reaction rate, the fission rate is affected more by the 1/r 2 effect. The reason is that both elastic and inelastic scattering end up with neutrons with energy above fission threshold, while only elastic scattering brings high energy neutrons to the sample and causes (n, 2n) reaction. A comparison is made with calculated results where the geometrical effects do not exist. (author)

  20. Spatial resolution of 2D ionization chamber arrays for IMRT dose verification: single-detector size and sampling step width

    International Nuclear Information System (INIS)

    Poppe, Bjoern; Djouguela, Armand; Blechschmidt, Arne; Willborn, Kay; Ruehmann, Antje; Harder, Dietrich

    2007-01-01

    The spatial resolution of 2D detector arrays equipped with ionization chambers or diodes, used for the dose verification of IMRT treatment plans, is limited by the size of the single detector and the centre-to-centre distance between the detectors. Optimization criteria with regard to these parameters have been developed by combining concepts of dosimetry and pattern analysis. The 2D-ARRAY Type 10024 (PTW-Freiburg, Germany), single-chamber cross section 5 x 5 mm 2 , centre-to-centre distance between chambers in each row and column 10 mm, served as an example. Additional frames of given dose distributions can be taken by shifting the whole array parallel or perpendicular to the MLC leaves by, e.g., 5 mm. The size of the single detector is characterized by its lateral response function, a trapezoid with 5 mm top width and 9 mm base width. Therefore, values measured with the 2D array are regarded as sample values from the convolution product of the accelerator generated dose distribution and this lateral response function. Consequently, the dose verification, e.g., by means of the gamma index, is performed by comparing the measured values of the 2D array with the values of the convolution product of the treatment planning system (TPS) calculated dose distribution and the single-detector lateral response function. Sufficiently small misalignments of the measured dose distributions in comparison with the calculated ones can be detected since the lateral response function is symmetric with respect to the centre of the chamber, and the change of dose gradients due to the convolution is sufficiently small. The sampling step width of the 2D array should provide a set of sample values representative of the sampled distribution, which is achieved if the highest spatial frequency contained in this function does not exceed the 'Nyquist frequency', one half of the sampling frequency. Since the convolution products of IMRT-typical dose distributions and the single-detector

  1. Novel shielding materials for space and air travel

    International Nuclear Information System (INIS)

    Vana, N.; Hajek, M.; Berger, T.; Fugger, M.; Hofmann, P.

    2006-01-01

    The reduction of dose onboard spacecraft and aircraft by appropriate shielding measures plays an essential role in the future development of space exploration and air travel. The design of novel shielding strategies and materials may involve hydrogenous composites, as it is well known that liquid hydrogen is most effective in attenuating charged particle radiation. As precursor for a later flight experiment, the shielding properties of newly developed hydrogen-rich polymers and rare earth-doped high-density rubber were tested in various ground-based neutron and heavy ion fields and compared with aluminium and polyethylene as reference materials. Absorbed dose, average linear energy transfer and gamma-equivalent neutron absorbed dose were determined by means of LiF:Mg,Ti thermoluminescence dosemeters and CR-39 plastic nuclear track detectors. First results for samples of equal aerial density indicate that selected hydrogen-rich plastics and rare-earth-doped rubber may be more effective in attenuating cosmic rays by up to 10% compared with conventional aluminium shielding. The appropriate adaptation of shielding thicknesses may thus allow reducing the biologically relevant dose. Owing to the lower density of the plastic composites, mass savings shall result in a significant reduction of launch costs. The experiment was flown as part of the European Space Agency's Biopan-5 mission in May 2005. (authors)

  2. Radiation shielding

    International Nuclear Information System (INIS)

    Yue, D.D.

    1979-01-01

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

  3. Shielded container

    International Nuclear Information System (INIS)

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  4. A feasibility study of a PET/MRI insert detector using strip-line and waveform sampling data acquisition.

    Science.gov (United States)

    Kim, H; Chen, C-T; Eclov, N; Ronzhin, A; Murat, P; Ramberg, E; Los, S; Wyrwicz, Alice M; Li, Limin; Kao, C-M

    2015-06-01

    We are developing a time-of-flight Positron Emission Tomography (PET) detector by using silicon photo-multipliers (SiPM) on a strip-line and high speed waveform sampling data acquisition. In this design, multiple SiPMs are connected on a single strip-line and signal waveforms on the strip-line are sampled at two ends of the strip to reduce readout channels while fully exploiting the fast time response of SiPMs. In addition to the deposited energy and time information, the position of the hit SiPM along the strip-line is determined by the arrival time difference of the waveform. Due to the insensitivity of the SiPMs to magnetic fields and the compact front-end electronics, the detector approach is highly attractive for developing a PET insert system for a magnetic resonance imaging (MRI) scanner to provide simultaneous PET/MR imaging. To investigate the feasibility, experimental tests using prototype detector modules have been conducted inside a 9.4 Tesla small animal MRI scanner (Bruker BioSpec 94/30 imaging spectrometer). On the prototype strip-line board, 16 SiPMs (5.2 mm pitch) are installed on two strip-lines and coupled to 2 × 8 LYSO scintillators (5.0 × 5.0 × 10.0 mm 3 with 5.2 mm pitch). The outputs of the strip-line boards are connected to a Domino-Ring-Sampler (DRS4) evaluation board for waveform sampling. Preliminary experimental results show that the effect of interference on the MRI image due to the PET detector is negligible and that PET detector performance is comparable with the results measured outside the MRI scanner.

  5. Influence of external and internal conditions of detector sample treatment on the particle registration sensitivity of Solid State Nuclear Track Detectors of type CR-39

    International Nuclear Information System (INIS)

    Hermsdorf, Dietrich

    2012-01-01

    The sensitivity of charged particle registration with SSNTD is the most important parameter to decide about the applicability of those detectors in research, technology and environmental dosimetry. The sensitivity is strongly influenced by the treatment of detector samples before, during and after the exposure and the final evaluation process by chemical etching. Whereas changes in detection properties by external environmental influences are generally considered, the dependences on the etching conditions are ignored. Commonly the sensitivity is assumed to compensate variations in the etching conditions for track revealing. In the present work the validity of this hypothesis will be checked. In the frame of the existing database the sensitivity is not really independent on variations in etching temperatures and should be corrected for differences in the activation energies for stimulation of the bulk and track etching process. Differences in the concentration dependence may be of minor importance. Furthermore, the registration sensitivity depends on environmental conditions before, during and after the irradiation with particles under investigation. Such external parameters are the air pressure, the sample temperature and modification of bulk material by out-gassing in vacuum and exposure to γ-rays. However, the available database is insufficient and inaccurate to draw final conclusions on the detection properties of SSNTD under various external and internal conditions.

  6. The scope of detector Medipix2 in micro-radiography of biological samples

    Czech Academy of Sciences Publication Activity Database

    Dammer, J.; Weyda, František; Jakůbek, J.; Škrabal, P.; Sopko, V.; Vavřík, D.

    2011-01-01

    Roč. 633, č. 1 (2011), s. 175-176 ISSN 0168-9002. [International Workshop on Radiation Imaging Detectors /11./. Praha, 29.06.2009-03.07.2009] R&D Projects: GA MŠk 2B06005 Grant - others:Ministerstvo školství(CZ) 6840770040; GA MŠk(CZ) 1P04LA211; GA MŠk(CZ) LC06041 Program:1P; LC Institutional research plan: CEZ:AV0Z50070508 Keywords : x-ray imaging * digital radiography * photon and x-ray detectors Subject RIV: EA - Cell Biology Impact factor: 1.207, year: 2011

  7. A new method for determining the uranium and thorium distribution in volcanic rock samples using solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Misdaq, M.A.; Bakhchi, A.; Ktata, A.; Koutit, A.; Lamine, J.; Ait nouh, F.; Oufni, L.

    2000-01-01

    A method based on using solid state nuclear track detectors (SSNTD) CR- 39 and LR-115 type II and calculating the probabilities for the alpha particles emitted by the uranium and thorium series to reach and be registered on these films was utilized for uranium and thorium contents determination in various geological samples. The distribution of uranium and thorium in different volcanic rocks has been investigated using the track fission method. In this work, the uranium and thorium contents have been determined in different volcanic rock samples by using CR-39 and LR-115 type II solid state nuclear track detectors (SSNTD). The mean critical angles of etching of the solid state nuclear track detectors utilized have been calculated. A petrographical study of the volcanic rock thin layers studied has been conducted. The uranium and thorium distribution inside different rock thin layers has been studied. The mechanism of inclusion of the uranium and thorium nuclei inside the volcanic rock samples studied has been investigated. (author)

  8. Quantitative portable gamma-spectroscopy sample analysis for non-standard sample geometries

    International Nuclear Information System (INIS)

    Ebara, S.B.

    1998-01-01

    Utilizing a portable spectroscopy system, a quantitative method for analysis of samples containing a mixture of fission and activation products in nonstandard geometries was developed. This method was not developed to replace other methods such as Monte Carlo or Discrete Ordinates but rather to offer an alternative rapid solution. The method can be used with various sample and shielding configurations where analysis on a laboratory based gamma-spectroscopy system is impractical. The portable gamma-spectroscopy method involves calibration of the detector and modeling of the sample and shielding to identify and quantify the radionuclides present in the sample. The method utilizes the intrinsic efficiency of the detector and the unattenuated gamma fluence rate at the detector surface per unit activity from the sample to calculate the nuclide activity and Minimum Detectable Activity (MDA). For a complex geometry, a computer code written for shielding applications (MICROSHIELD) is utilized to determine the unattenuated gamma fluence rate per unit activity at the detector surface. Lastly, the method is only applicable to nuclides which emit gamma-rays and cannot be used for pure beta or alpha emitters. In addition, if sample self absorption and shielding is significant, the attenuation will result in high MDA's for nuclides which solely emit low energy gamma-rays. The following presents the analysis technique and presents verification results using actual experimental data, rather than comparisons to other approximations such as Monte Carlo techniques, to demonstrate the accuracy of the method given a known geometry and source term. (author)

  9. Quantitative portable gamma spectroscopy sample analysis for non-standard sample geometries

    International Nuclear Information System (INIS)

    Enghauser, M.W.; Ebara, S.B.

    1997-01-01

    Utilizing a portable spectroscopy system, a quantitative method for analysis of samples containing a mixture of fission and activation products in nonstandard geometries was developed. The method can be used with various sample and shielding configurations where analysis on a laboratory based gamma spectroscopy system is impractical. The portable gamma spectroscopy method involves calibration of the detector and modeling of the sample and shielding to identify and quantify the radionuclides present in the sample. The method utilizes the intrinsic efficiency of the detector and the unattenuated gamma fluence rate at the detector surface per unit activity from the sample to calculate the nuclide activity and Minimum Detectable Activity (MDA). For a complex geometry, a computer code written for shielding applications (MICROSHIELD) is utilized to determine the unattenuated gamma fluence rate per unit activity at the detector surface. Lastly, the method is only applicable to nuclides which emit gamma rays and cannot be used for pure beta emitters. In addition, if sample self absorption and shielding is significant, the attenuation will result in high MDA's for nuclides which solely emit low energy gamma rays. The following presents the analysis technique and presents verification results demonstrating the accuracy of the method

  10. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Directory of Open Access Journals (Sweden)

    Hegazy Aya Hamdy

    2018-01-01

    Full Text Available Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1 shielding-collimator material, (2 Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3 thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  11. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Science.gov (United States)

    Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.

    2018-04-01

    Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  12. Development of sample assay system equipped with 3He Alternative Neutron Detectors (ASAS). (2) Results of ASAS measurement test

    International Nuclear Information System (INIS)

    Tanigawa, Masafumi; Mukai, Yasunobu; Kurita, Tsutomu; Makino, Risa; Nakamura, Hironobu; Tobita, Hiroshi; Ohzu, Akira; Kureta, Masatoshi; Seya, Michio

    2015-01-01

    Against the background of the serious shortage of 3 He gas, design and development of a new detector equipped ZnS/ 10 B 2 O 3 ceramic scintillation neutron detectors in JAEA, with the support of the government (the Ministry of Education, Culture, Sports, Science and Technology). The design of the alternative 3 He detector is referred from INVS (INVentory Sample assay system (HLNCC (High Level Neutron Coincidence Counter) type)) which is being used for the verification of MOX powder etc. and is named it as ASAS (Alternative Sample Assay System). In order to prove the Pu quantitative performance as an alternative technology, several measurement tests and comparison test with INVS were conducted using ASAS. In these tests, evaluation of fundamental performance (counting efficiency and die-away time) and uncertainty evaluations were implemented. As a result, although fundamental performance of ASAS was not achieved to the one of INVS, we could confirm that ASAS has almost the same Pu quantitative performance including measurement uncertainty as that of INVS. (author)

  13. Effect of detector collimator and sample thickness on 0.662 MeV multiply Compton-scattered gamma rays

    International Nuclear Information System (INIS)

    Singh, Manpreet; Singh, Gurvinderjit; Sandhu, B.S.; Singh, Bhajan

    2006-01-01

    The simultaneous effect of detector collimator and sample thickness on 0.662 MeV multiply Compton-scattered gamma photons was studied experimentally. An intense collimated beam, obtained from 6-Ci 137 Cs source, is allowed to impinge on cylindrical aluminium samples of varying diameter and the scattered photons are detected by a 51 mmx51 mm NaI(Tl) scintillation detector placed at 90 o to the incident beam. The full energy peak corresponding to singly scattered events is reconstructed analytically. The thickness at which the multiply scattered events saturate is determined for different detector collimators. The parameters like signal-to-noise ratio and multiply scatter fraction (MSF) have also been deduced and support the work carried out by Shengli et al. [2000. EGS4 simulation of Compton scattering for nondestructive testing. KEK proceedings 200-20, Tsukuba, Japan, pp. 216-223] and Barnea et al. [1995. A study of multiple scattering background in Compton scatter imaging. NDT and E International 28, 155-162] based upon Monte Carlo calculations

  14. Hydrogen, carbon and oxygen determination in proxy material samples using a LaBr3:Ce detector.

    Science.gov (United States)

    Naqvi, A A; Al-Matouq, Faris A; Khiari, F Z; Isab, A A; Raashid, M; Khateeb-ur-Rehman

    2013-08-01

    Hydrogen, carbon and oxygen concentrations were measured in caffeine, urea, ammonium acetate and melamine bulk samples via 14 MeV neutron inelastic scattering using a LaBr3:Ce detector. The samples tested herein represent drugs, explosives and benign materials, respectively. Despite its intrinsic activity, the LaBr3:Ce detector performed well in detecting the hydrogen, carbon and oxygen elements. Because 5.1 MeV nitrogen gamma rays interfere with silicon and calcium prompt gamma rays from the room background, the nitrogen peak was not detected in the samples. An excellent agreement was observed between the experimental and theoretical yields of 2.22, 4.43 and 6.13 MeV gamma rays from the analyzed samples as a function of H, C and O concentrations, respectively. Within statistical errors, the minimum detectable concentration (MDC) of hydrogen, carbon and oxygen elements in the tested materials were consistent with previously reported MDC values for these elements measured in hydrocarbon samples. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. Particle identification using digital pulse shape discrimination in a nTD silicon detector with a 1 GHz sampling digitizer

    Science.gov (United States)

    Mahata, K.; Shrivastava, A.; Gore, J. A.; Pandit, S. K.; Parkar, V. V.; Ramachandran, K.; Kumar, A.; Gupta, S.; Patale, P.

    2018-06-01

    In beam test experiments have been carried out for particle identification using digital pulse shape analysis in a 500 μm thick Neutron Transmutation Doped (nTD) silicon detector with an indigenously developed FPGA based 12 bit resolution, 1 GHz sampling digitizer. The nTD Si detector was used in a low-field injection setup to detect light heavy-ions produced in reactions of ∼ 5 MeV/A 7Li and 12C beams on different targets. Pulse height, rise time and current maximum have been obtained from the digitized charge output of a high bandwidth charge and current sensitive pre-amplifier. Good isotopic separation have been achieved using only the digitized charge output in case of light heavy-ions. The setup can be used for charged particle spectroscopy in nuclear reactions involving light heavy-ions around the Coulomb barrier energies.

  16. New sampling electronics using CCD for DIOGENE: a high multiplicity, 4 π detector for relativistic heavy ions

    International Nuclear Information System (INIS)

    Babinet, R.P.

    1987-01-01

    DIOGENE is a small time projection chamber which has been developed to study central collisions of relativistic heavy ions. The maximum multiplicity (up to 40 charged particles) that can be accepted by this detector is limited by the present electronics. In view of the heavier mass ions that should become readily available at the Saturne national facility (France), a new sampling electronics has been tested. In the first part of this talk they will present a brief description of the actual detector, insisting on the performances that have been effectively obtained with α-particles and Neon beams. The motivation for and characteristics of a renewed electronic set-up should thus appear more clearly. The second part of the talk is devoted to results of the tests that have been performed using charged couple devices. They will finally conclude on the future perspectives that have been opened by these developments

  17. Activation of TRIGA Mark II research reactor concrete shield

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz; Bozic, Matjaz

    2002-01-01

    To determine neutron activation inside the TRIGA research reactor concrete body a special sample-holder for irradiation inside horizontal channel was developed and tested. In the sample-holder various samples can be irradiated at different concrete shielding depths. In this paper the description of the sample-holder, experiment conditions and results of long-lived activation measurements are given. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active long-lived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale. (author)

  18. Radiation shielding material

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. Optical properties studies of glass samples for prototyping a TORCH detector module

    CERN Multimedia

    Castillo García, L

    2014-01-01

    TORCH (Time Of internally Reflected CHerenkov light) ) is a proposed particle identification system to achieve positive π/K/p separation at a ≥3σ level in the momentum range below 10 GeV/c. Cherenkov photons are generated from charged particle tracks crossing a 1cm-thick quartz plate. They propagate by total internal reflection to the edge and are focused onto an array of micro-channel plate photon detectors. Their position and arrival time are recorded. This allows the reconstruction of the photon trajectory and the particle crossing time. Results on optical tests are presented.

  20. Variable sampling-time technique for improving count rate performance of scintillation detectors

    International Nuclear Information System (INIS)

    Tanaka, E.; Nohara, N.; Murayama, H.

    1979-01-01

    A new technique is presented to improve the count rate capability of a scintillation spectrometer or a position sensitive detector with minimum loss of resolution. The technique is based on the combination of pulse shortening and selective integration in which the integration period is not fixed but shortened by the arrival of the following pulse. Theoretical analysis of the degradation of the statiscal component of resolution is made for the proposed system with delay line pulse shortening, and the factor of resolution loss is formulated as a function of the input pulse rate. A new method is also presented for determining the statistical component of resolution separately from the non-statistical system resolution. Preliminary experiments with a NaI(Tl) detector have been carried out, the results of which are consistent with the theoretical prediction. However, due to the non-exponential scintillation decay of the NaI(Tl) crystal, a simple delay line clipping is not satisfactory, and an RC high-pass filter has been added, which results in further degradation of the statistical resolution. (Auth.)

  1. Study of sample-detector assemblies for application to in-situ measurement of radioactivity in liquid effluents

    International Nuclear Information System (INIS)

    Pendharkar, K.A.; Narayanan Kutty, K.; Krishnamony, S.

    1991-01-01

    This paper describes the experimental investigations carried out on four different types of sample-detector assemblies with a view to determining their detection limits and relative merits for application to in-situ measurement of radioactivity in liquid effluents. The four systems studied were: (1) gamma detection using 11 cm x 8 cm NaI (Tl) scintillation detector inserted in the cavity of a specially designed stainless steel chamber of capacity 15 liters, (2) gamma detection using a metal-walled G.M. counter in a similar manner, (3) beta detection using twin thin-walled G.M. counters immersed in liquid, and (4) end window G.M. counter positioned above the liquid surface in a shallow tray. The design features of an in-line monitor employing a 11 cm x 8 cm NaI (Tl) detector used for the routine monitoring of beta gamma activity concentrations in the low level effluents of the Tarapur Fuel Processing Plant are described. (author). 1 tab

  2. Measuring space radiation shielding effectiveness

    Directory of Open Access Journals (Sweden)

    Bahadori Amir

    2017-01-01

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

  3. Measuring space radiation shielding effectiveness

    Science.gov (United States)

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

    2017-09-01

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

  4. About the Scythian Shields

    Directory of Open Access Journals (Sweden)

    About the Scythian Shields

    2017-10-01

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

  5. Gamma ray shielding: a web based interactive program

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Mobile robot prototype detector of gamma radiation

    International Nuclear Information System (INIS)

    Vazquez C, R.M.; Duran V, M. D.; Jardon M, C. I.

    2014-10-01

    In this paper the technological development of a mobile robot prototype detector of gamma radiation is shown. This prototype has been developed for the purpose of algorithms implementation for the applications of terrestrial radiation monitoring of exposed sources, search for missing radioactive sources, identification and delineation of radioactive contamination areas and distribution maps generating of radioactive exposure. Mobile robot detector of radiation is an experimental technology development platform to operate in laboratory environment or flat floor facilities. The prototype integrates a driving section of differential configuration robot on wheels, a support mechanism and rotation of shielded detector, actuator controller cards, acquisition and processing of sensor data, detection algorithms programming and control actuators, data recording (Data Logger) and data transmission in wireless way. The robot in this first phase is remotely operated in wireless way with a range of approximately 150 m line of sight and can extend that range to 300 m or more with the use of signal repeaters. The gamma radiation detection is performed using a Geiger detector shielded. Scan detection is performed at various time sampling periods and diverse positions of discrete or continuous angular orientation on the horizon. The captured data are geographical coordinates of robot GPS (latitude and longitude), orientation angle of shield, counting by sampling time, date, hours, minutes and seconds. The data is saved in a file in the Micro Sd memory on the robot. They are also sent in wireless way by an X Bee card to a remote station that receives for their online monitoring on a laptop through an acquisition program by serial port on Mat Lab. Additionally a voice synthesizing card with a horn, both in the robot, periodically pronounced in Spanish, data length, latitude, orientation angle of shield and detected accounts. (Author)

  7. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray...

  8. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — We utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray spectrometers. Two...

  9. Shielding benchmark problems

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  10. Multicounter neutron detector for examination of content and spatial distribution of fissile materials in bulk samples

    International Nuclear Information System (INIS)

    Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

    1999-01-01

    A new neutron coincidence well-counter is presented. This experimental device can be applied for passive assay of fissile and, in particular, for plutonium bearing materials. It contains of a set of the 3 He tubes placed inside a polyethylene moderator. Outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using a correlator connected with PC, and correlation techniques implemented in software. Such a neutron counter enables determination of the 240 Pu effective mass in samples of a small Pu content (i.e., where the multiplication effects can be neglected) having a fairly big volume (up to 0.17 m 3 ), if only the isotopic composition is known. For determination of neutron sources distribution inside a sample, a heuristic method based on hierarchical cluster analysis was applied. As input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples were taken. Such matrices of profiles counts are collected using the sample scanning with detection head. In the clustering processes, process, counts profiles of unknown samples are fitted into dendrograms employing the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in the examined sample is then evaluated on the basis of a comparison with standard sources distributions. (author)

  11. Radiation shielding device

    International Nuclear Information System (INIS)

    Nakagawa, Takahiro; Yamagami, Makoto.

    1996-01-01

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

  12. Determination of gamma emitting radionuclides in environmental air and precipitation samples with a Ge(Li) detector

    International Nuclear Information System (INIS)

    Hoetzl, H.; Rosner, G.; Winkler, R.; Sansoni, B.

    1977-01-01

    The concentrations of the radionuclides 7 Be, 54 Mn, 95 Zr, 95 Nb, 103 Ru, 106 Ru, 125 Sb, 137 Cs, 140 Ba/ 140 La, 141 Ce and 144 Ce in ground level air and of 7 Be, 95 Zr, 137 Cs and 144 Ce in precipitation were determined since 1970 and 1971 respectively at Neuherberg, 10 km north of Munich, by gamma spectrometry using a 60 cm 3 Ge(Li) detector. Dust samples were collected twice a month 1 m above ground from about 40,000 m 3 of air on 46 cm x 28 cm microsorbane filters and pressed to small cylinders of 35 cm 3 in size. Sensitivity of the procedure is of the order of 1 fCi/m 3 for air and of 10 pCi/m 2 per month for precipitation samples at a counting time of 1500 min. (author)

  13. 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...

  14. The TPC shielding of the CAST experiment

    International Nuclear Information System (INIS)

    Ruz, J; Luzon, G; Beltran, B; Carmona, J M; Cebrian, S; Gomez, H; Irastorza, I G; Morales, J; Ortiz de Solorzano, A; RodrIguez, A; Villar, J A

    2006-01-01

    Sunset solar axions traversing the intense magnetic field of the CERN Axion Solar Telescope (CAST) experiment may be detected in a TPC detector, placed at one side of the magnet, as point-like X-rays signals. This signal could be masked, however, by the inhomogeneous radioactive background of materials and experimental site. Here we present the shielding built to reduce and homogenize the radioactive background levels of the TPC detector

  15. A simulation study of high-resolution x-ray computed tomography imaging using irregular sampling with a photon-counting detector

    International Nuclear Information System (INIS)

    Lee, Seungwan; Choi, Yu-Na; Kim, Hee-Joung

    2013-01-01

    The purpose of this study was to improve the spatial resolution for the x-ray computed tomography (CT) imaging with a photon-counting detector using an irregular sampling method. The geometric shift-model of detector was proposed to produce the irregular sampling pattern and increase the number of samplings in the radial direction. The conventional micro-x-ray CT system and the novel system with the geometric shift-model of detector were simulated using analytic and Monte Carlo simulations. The projections were reconstructed using filtered back-projection (FBP), algebraic reconstruction technique (ART), and total variation (TV) minimization algorithms, and the reconstructed images were compared in terms of normalized root-mean-square error (NRMSE), full-width at half-maximum (FWHM), and coefficient-of-variation (COV). The results showed that the image quality improved in the novel system with the geometric shift-model of detector, and the NRMSE, FWHM, and COV were lower for the images reconstructed using the TV minimization technique in the novel system with the geometric shift-model of detector. The irregular sampling method produced by the geometric shift-model of detector can improve the spatial resolution and reduce artifacts and noise for reconstructed images obtained from an x-ray CT system with a photon-counting detector. -- Highlights: • We proposed a novel sampling method based on a spiral pattern to improve the spatial resolution. • The novel sampling method increased the number of samplings in the radial direction. • The spatial resolution was improved by the novel sampling method

  16. MARMER, a flexible point-kernel shielding code

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Hoogenboom, J.E.

    1990-01-01

    A point-kernel shielding code entitled MARMER is described. It has several options with respect to geometry input, source description and detector point description which extend the flexibility and usefulness of the code, and which are especially useful in spent fuel shielding. MARMER has been validated using the TN12 spent fuel shipping cask benchmark. (author)

  17. MARMER, a flexible point-kernel shielding code

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1990-01-01

    A point-kernel shielding code entitled MARMER is described. It has several options with respect to geometry input, source description and detector point description which extend the flexibility and usefulness of the code, and which are especially useful in spent fuel shielding. MARMER has been validated using the TN12 spent fuel shipping cask benchmark. (author).

  18. Application of the Monte Carlo method for the efficiency calibration of CsI and NaI detectors for gamma-ray measurements from terrestrial samples

    International Nuclear Information System (INIS)

    Baccouche, S.; Al-Azmi, D.; Karunakara, N.; Trabelsi, A.

    2012-01-01

    Gamma-ray measurements in terrestrial/environmental samples require the use of high efficient detectors because of the low level of the radionuclide activity concentrations in the samples; thus scintillators are suitable for this purpose. Two scintillation detectors were studied in this work; CsI(Tl) and NaI(Tl) with identical size for measurement of terrestrial samples for performance study. This work describes a Monte Carlo method for making the full-energy efficiency calibration curves for both detectors using gamma-ray energies associated with the decay of naturally occurring radionuclides 137 Cs (661 keV), 40 K (1460 keV), 238 U ( 214 Bi, 1764 keV) and 232 Th ( 208 Tl, 2614 keV), which are found in terrestrial samples. The magnitude of the coincidence summing effect occurring for the 2614 keV emission of 208 Tl is assessed by simulation. The method provides an efficient tool to make the full-energy efficiency calibration curve for scintillation detectors for any samples geometry and volume in order to determine accurate activity concentrations in terrestrial samples. - Highlights: ► CsI (Tl) and NaI (Tl) detectors were studied for the measurement of terrestrial samples. ► Monte Carlo method was used for efficiency calibration using natural gamma emitting terrestrial radionuclides. ► The coincidence summing effect occurring for the 2614 keV emission of 208 Tl is assessed by simulation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  20. Noise, sampling, and the number of projections in cone-beam CT with a flat-panel detector

    International Nuclear Information System (INIS)

    Zhao, Z.; Gang, G. J.; Siewerdsen, J. H.

    2014-01-01

    Purpose: To investigate the effect of the number of projection views on image noise in cone-beam CT (CBCT) with a flat-panel detector. Methods: This fairly fundamental consideration in CBCT system design and operation was addressed experimentally (using a phantom presenting a uniform medium as well as statistically motivated “clutter”) and theoretically (using a cascaded systems model describing CBCT noise) to elucidate the contributing factors of quantum noise (σ Q ), electronic noise (σ E ), and view aliasing (σ view ). Analysis included investigation of the noise, noise-power spectrum, and modulation transfer function as a function of the number of projections (N proj ), dose (D tot ), and voxel size (b vox ). Results: The results reveal a nonmonotonic relationship between image noise andN proj at fixed total dose: for the CBCT system considered, noise decreased with increasing N proj due to reduction of view sampling effects in the regime N proj proj due to increased electronic noise. View sampling effects were shown to depend on the heterogeneity of the object in a direct analytical relationship to power-law anatomical clutter of the form κ/f  β —and a general model of individual noise components (σ Q , σ E , and σ view ) demonstrated agreement with measurements over a broad range in N proj , D tot , and b vox . Conclusions: The work elucidates fairly basic elements of CBCT noise in a manner that demonstrates the role of distinct noise components (viz., quantum, electronic, and view sampling noise). For configurations fairly typical of CBCT with a flat-panel detector (FPD), the analysis reveals a “sweet spot” (i.e., minimum noise) in the rangeN proj ∼ 250–350, nearly an order of magnitude lower in N proj than typical of multidetector CT, owing to the relatively high electronic noise in FPDs. The analysis explicitly relates view aliasing and quantum noise in a manner that includes aspects of the object (“clutter”) and imaging chain

  1. Energy-dispersive X-ray fluorescence of discarded tire samples, using a Si-PIN detector

    International Nuclear Information System (INIS)

    Lopes, Fabio; Appoloni, C.R.; Melquiades, Fabio L.

    2007-01-01

    The determination of zinc concentration in samples of discarded tires is of great environmental interest because the process for manufacturing tyres uses S for rubber vulcanization, and ZnO is the reaction catalyst. Discarded tyres are being used in asphalt paving, in the burning process of thermoelectric and cement industries and also for controlling erosion in agricultural areas. Analysis of tyre samples usually requires chemical digestion which is slow and expensive. Aiming to eliminate those limitations, this work uses energy-dispersive X-ray fluorescence (EDXRF) with a portable equipment, once it is a simultaneous multi-element analytical technique, requiring minimal sample preparation. Five samples of discarded tyres have been ground and analysed in the form of pastilles, using a mini X-ray tube (Ag target, MO filter, 25 kV/20 μA) for 200 s, and a Si-PIN semiconductor detector coupled to a multichannel analyser. Zinc concentrations in the range of 40.6 to 44.2 μg g -1 have been obtained, representing 0.4% of the tire composition, which is below the maximum value (2%) recommended by the European Tyre Recycling Association. Concentrations between 0.15 and 0.52 μg g -1 were obtained for Fe

  2. Mobile robot prototype detector of gamma radiation; Prototipo de robot movil detector de radiacion gamma

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez C, R.M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Duran V, M. D.; Jardon M, C. I., E-mail: raulmario.vazquez@inin.gob.mx [Tecnologico de Estudios Superiores de Villa Guerrero, Carretera Federal Toluca-Ixtapan de la Sal Km. 64.5, La Finca Villa Guerrero, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the technological development of a mobile robot prototype detector of gamma radiation is shown. This prototype has been developed for the purpose of algorithms implementation for the applications of terrestrial radiation monitoring of exposed sources, search for missing radioactive sources, identification and delineation of radioactive contamination areas and distribution maps generating of radioactive exposure. Mobile robot detector of radiation is an experimental technology development platform to operate in laboratory environment or flat floor facilities. The prototype integrates a driving section of differential configuration robot on wheels, a support mechanism and rotation of shielded detector, actuator controller cards, acquisition and processing of sensor data, detection algorithms programming and control actuators, data recording (Data Logger) and data transmission in wireless way. The robot in this first phase is remotely operated in wireless way with a range of approximately 150 m line of sight and can extend that range to 300 m or more with the use of signal repeaters. The gamma radiation detection is performed using a Geiger detector shielded. Scan detection is performed at various time sampling periods and diverse positions of discrete or continuous angular orientation on the horizon. The captured data are geographical coordinates of robot GPS (latitude and longitude), orientation angle of shield, counting by sampling time, date, hours, minutes and seconds. The data is saved in a file in the Micro Sd memory on the robot. They are also sent in wireless way by an X Bee card to a remote station that receives for their online monitoring on a laptop through an acquisition program by serial port on Mat Lab. Additionally a voice synthesizing card with a horn, both in the robot, periodically pronounced in Spanish, data length, latitude, orientation angle of shield and detected accounts. (Author)

  3. Handout on shielding calculation

    International Nuclear Information System (INIS)

    Heilbron Filho, P.F.L.

    1991-01-01

    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)

  4. Study of /sup 210/Pb and /sup 210/Po distributions in environmental samples by CR-39 track detector

    Energy Technology Data Exchange (ETDEWEB)

    Hunyadi, I.; Somogyi, G.; Szilagyi, S. (Magyar Tudomanyos Akademia, Debrecen. Atommag Kutato Intezete)

    1984-01-01

    Activity concentration distributions of long-lived alpha-emitters in aerosol samples are analysed by high-resolution autoradiography in CR-39. A study of the alpha-activity attached to aerosols of different particulate sizes separated by a cascade impactor is also performed. It is found that, in the majority of samples, the alpha-activity can be dominantly related to the presence of /sup 210/Po produced by its beta-active precursor /sup 210/Pb. In our studies we have applied the following methods: 1) analysis of alpha-decay properties by means of autoradiographs taken at different post-sampling times, 2) spectroscopical study of individual alpha-tracks and track clusters by a method developed by us for high-resolution alpha-energy determination. In the second method the parameters to be measured are the major axis of surface track opening, the diameter of etched out track end, the total length measurable on the surface along the projected track, and the thickness of layer etched away from the detector surface.

  5. Design of emergency shield

    International Nuclear Information System (INIS)

    Soliman, S.E.

    1993-01-01

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

  6. Shielding effectiveness of superconductive particles in plastics

    International Nuclear Information System (INIS)

    Pienkowski, T.; Kincaid, J.; Lanagan, M.T.; Poeppel, R.B.; Dusek, J.T.; Shi, D.; Goretta, K.C.

    1988-09-01

    The ability to cool superconductors with liquid nitrogen instead of liquid helium has opened the door to a wide range of research. The well known Meissner effect, which states superconductors are perfectly diamagnetic, suggests shielding applications. One of the drawbacks to the new ceramic superconductors is the brittleness of the finished material. Because of this drawback, any application which required flexibility (e.g., wire and cable) would be impractical. Therefore, this paper presents the results of a preliminary investigation into the shielding effectiveness of YBa 2 Cu 3 O/sub 7-x/ both as a composite and as a monolithic material. Shielding effectiveness was measured using two separate test methods. One tested the magnetic (near field) shielding, and the other tested the electromagnetic (far field) shielding. No shielding was seen in the near field measurements on the composite samples, and only one heavily loaded sample showed some shielding in the far field. The monolithic samples showed a large amount of magnetic shielding. 5 refs., 5 figs

  7. Electromagnetically shielded building

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Electromagnetically shielded building

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-21

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

  9. Revise of a basic data base for shielding design

    International Nuclear Information System (INIS)

    Nakao, Makoto; Takemura, Morio

    2000-03-01

    With use of the two-dimensional discrete ordinates code DORT and the standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the representative configurations in the Radial Shield Attenuation Experiment of the JASPER were performed. The results were compared with those obtained with use of traditional method DOT3.5/JSDJ2 for the previous JASPER experimental analyses. In general, the change of the cross section library gives higher results and the change of the transport code gives lower results. Finally the new analysis method gives better agreement with the experimental results and also less deviations of calculational errors between various detectors. Experimental analyses for the thick concrete configuration in the Gap Streaming Experiment of the JASPER was also performed with the new analysis method, after solving the poor agreement found in last year with the original JASPER experimental analyses. The same tendency due to the library change was confirmed with the above mentioned analyses of the Radial Shield Attenuation Experiment. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments were continued through the above-mentioned experimental analyses and related informations were added for repletion of the database preserved in a computer disk holding previously accumulated data. Input data descriptions were made for auxiliary routines needed for the experimental analyses and their sample data were compiled and stored in the database. (author)

  10. Test and performance of a BGO Compton-suppression shield for GAMMASPHERE

    International Nuclear Information System (INIS)

    Carpenter, M.P.; Khoo, T.L.; Ahmad, I.

    1994-01-01

    Bismuth germanate (BGO) compton-suppression shields have been constructed to surround the Ge detectors of the GAMMASPHERE array. A shield consists of six hexagonal tapered BGO elements, each coupled to two 1-inch x 1-inch photomultiplier tubes. In addition, a cylindrical BGO detector is placed behind the Ge detector to intercept the forward scattered gamma rays. One hundred ten such shields are planned for the GAMMASPHERE array. Procedures for measuring the performance of these shields have been developed. Large (70 %) Ge detectors when used with these shields give a peak-to-total ratio of better tan 0.60. To date more than 85 shield have been tested and approved for use in GAMMASPHERE

  11. Electromagnetic shielding formulae

    International Nuclear Information System (INIS)

    Dahlberg, E.

    1979-02-01

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

  12. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  13. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

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

  14. Split detector

    International Nuclear Information System (INIS)

    Cederstrand, C.N.; Chism, H.R.

    1982-01-01

    A gas analyzer is disclosed which provides a dual channel capability for the simultaneous determination of the presence and concentration of two gases in a stream of sample gas and which has a single infrared source, a single sample cell, two infrared bandpass filters, and two infrared detectors. A separator between the filters and detectors prevents interchange of radiation between the filters. The separator is positioned by fitting it in a slot

  15. 32P-postlabeling assay for carcinogen-DNA adducts: description of beta shielding apparatus and semi-automatic spotting and washing devices that facilitate the handling of multiple samples

    International Nuclear Information System (INIS)

    Reddy, M.V.; Blackburn, G.R.

    1990-01-01

    The utilization of the 32 P-postlabeling assay in combination with TLC for the sensitive detection and estimation of aromatic DNA adducts has been increasing. The procedure consists of 32 P-labeling of carcinogen-adducted 3'-nucleotides in the DNA digests using γ- 32 P ATP and polynucleotide kinase, separation of 32 P-labeled adducts by TLC, and their detection by autoradiography. During both 32 P-labeling and initial phases of TLC, a relatively high amount of γ- 32 P ATP is handled when 30 samples are processed simultaneously. We describe the design of acrylic shielding apparatus, semi-automatic TLC spotting devices, and devices for development and washing of multiple TLC plates, which not only provide substantial protection from exposure to 32 P beta radiation, but also allow quick and easy handling of a large number of samples. Specifically, the equipment includes: (i) a multi-tube carousel rack having 15 wells to hold capless Eppendorf tubes and a rotatable lid with an aperture to access individual tubes; (ii) a pipette shielder; (iii) two semi-automatic spotting devices to apply radioactive solutions to TLC plates; (iv) a multi-plate holder for TLC plates; and (v) a mechanical device for washing multiple TLC plates. Item (i) is small enough to be held in one-hand, vortexed, and centrifuged to mix the solutions in each tube while beta radiation is shielded. Items (iii) to (iv) aid in the automation of the assay. (author)

  16. Application of the Monte Carlo method for the efficiency calibration of CsI and NaI detectors for gamma-ray measurements from terrestrial samples

    Energy Technology Data Exchange (ETDEWEB)

    Baccouche, S., E-mail: souad.baccouche@cnstn.rnrt.tn [UR-MDTN, National Center for Nuclear Sciences and Technology, Technopole Sidi Thabet, 2020 Sidi Thabet (Tunisia); Al-Azmi, D., E-mail: ds.alazmi@paaet.edu.kw [Department of Applied Sciences, College of Technological Studies, Public Authority for Applied Education and Training, Shuwaikh, P.O. Box 42325, Code 70654 (Kuwait); Karunakara, N., E-mail: karunakara_n@yahoo.com [University Science Instrumentation Centre, Mangalore University, Mangalagangotri 574199 (India); Trabelsi, A., E-mail: adel.trabelsi@fst.rnu.tn [UR-MDTN, National Center for Nuclear Sciences and Technology, Technopole Sidi Thabet, 2020 Sidi Thabet (Tunisia); UR-UPNHE, Faculty of Sciences of Tunis, El-Manar University, 2092 Tunis (Tunisia)

    2012-01-15

    Gamma-ray measurements in terrestrial/environmental samples require the use of high efficient detectors because of the low level of the radionuclide activity concentrations in the samples; thus scintillators are suitable for this purpose. Two scintillation detectors were studied in this work; CsI(Tl) and NaI(Tl) with identical size for measurement of terrestrial samples for performance study. This work describes a Monte Carlo method for making the full-energy efficiency calibration curves for both detectors using gamma-ray energies associated with the decay of naturally occurring radionuclides {sup 137}Cs (661 keV), {sup 40}K (1460 keV), {sup 238}U ({sup 214}Bi, 1764 keV) and {sup 232}Th ({sup 208}Tl, 2614 keV), which are found in terrestrial samples. The magnitude of the coincidence summing effect occurring for the 2614 keV emission of {sup 208}Tl is assessed by simulation. The method provides an efficient tool to make the full-energy efficiency calibration curve for scintillation detectors for any samples geometry and volume in order to determine accurate activity concentrations in terrestrial samples. - Highlights: Black-Right-Pointing-Pointer CsI (Tl) and NaI (Tl) detectors were studied for the measurement of terrestrial samples. Black-Right-Pointing-Pointer Monte Carlo method was used for efficiency calibration using natural gamma emitting terrestrial radionuclides. Black-Right-Pointing-Pointer The coincidence summing effect occurring for the 2614 keV emission of {sup 208}Tl is assessed by simulation.

  17. Wake Shield Target Protection

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  18. Evaluation of a gas chromatograph with a novel surface acoustic wave detector (SAW GC) for screening of volatile organic compounds in Hanford waste tank samples

    International Nuclear Information System (INIS)

    Lockrem, L.L.

    1998-01-01

    A novel instrument, a gas chromatograph with a Surface Acoustic Wave Detector (SAW GC), was evaluated for the screening of organic compounds in Hanford tank headspace vapors. Calibration data were developed for the most common organic compounds, and the accuracy and precision were measured with a certified standard. The instrument was tested with headspace samples collected from seven Hanford waste tanks

  19. Radiation shielding concrete

    International Nuclear Information System (INIS)

    Kunishima, Shigeru.

    1990-01-01

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

  20. Accelerator shielding benchmark problems

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Development and simulation of a Ge/Si multi-detector spectrometer for fission products traces detection in the environment

    International Nuclear Information System (INIS)

    Cagniant, Antoine

    2015-01-01

    For the verification of the Comprehensive nuclear Test Ban Treaty (CTBT), the measurement of fission products trace levels in the environment is fundamental. Such measurement is a key indicator of a nuclear explosion. For constant amelioration of these measurements, the CEA/DAM-Ile de France has developed and installed a new dedicated surface spectrometer. Named GAMMA3, it is equipped with three germanium detectors, two silicon detectors (integrated in a dedicated gas cell, the PIPSBox) and includes an optimized shielding.This shielding reduces greatly the interference of environmental photons, muons and neutrons with the detectors. The residual radiological background measured inside the shielding is the community's lowest for a surface laboratory. This set of high energy resolution detectors allows the operator to optimize a measurement according to the sample geometry, activity or nature. More precisely, a radioactive noble gas can be measured by photon/electron coincidence, an active sample can be measured by photon/photon coincidence, and a low-active sample can be measured in a high-efficiency configuration. Combining optimized shielding and optimized measurement, Minimum Detectable Activities required for CTBT certification are obtained quickly. Specifically, MDA is reached in 5 hours for 140-Ba (24 mBq), in 6h30 hours for 131m/133m-Xe (5 mBq) and in 7h15 for 133-Xe (5 mBq), when CTBT requirement is in 6 days. (author) [fr

  2. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

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

  3. Seismic proof test of shielding block walls

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. Radiation shielding plate

    International Nuclear Information System (INIS)

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

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

  5. Shielding design for positron emission tomography facility

    International Nuclear Information System (INIS)

    Abdallah, I.I.

    2007-01-01

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

  6. Shielding correction to bodywork of in-situ object counting system

    International Nuclear Information System (INIS)

    Feng Tiancheng; Chen Wei; Long Bin; Su Chuanying; Wu Rui; Jia Mingyan; Cheng Jianping

    2009-01-01

    This paper presents the methods of experiment and calculation for shielding correction to the bodywork of in-situ object counting system (ISOCS) using a plane source of 152 Eu. The shielding correction coefficients were obtained in the conditions that the HPGe detector of BE5030 with the collimators of 50 mm-90 degree, 50 mm-30 degree or 50 mm-180 degree, and the detector distance 58.2 cm from ground surface. The relationships between the shielding correction coefficients and γ-ray energies were fitted by the least square method, for the shielding correction calculation of any energy within 122-1 408 keV by interpolation. (authors)

  7. The use of laser-induced fluorescence or ultraviolet detectors for sensitive and selective analysis of tobramycin or erythropoietin in complex samples

    Science.gov (United States)

    Ahmed, Hytham M.; Ebeid, Wael B.

    2015-05-01

    Complex samples analysis is a challenge in pharmaceutical and biopharmaceutical analysis. In this work, tobramycin (TOB) analysis in human urine samples and recombinant human erythropoietin (rhEPO) analysis in the presence of similar protein were selected as representative examples of such samples analysis. Assays of TOB in urine samples are difficult because of poor detectability. Therefore laser induced fluorescence detector (LIF) was combined with a separation technique, micellar electrokinetic chromatography (MEKC), to determine TOB through derivatization with fluorescein isothiocyanate (FITC). Borate was used as background electrolyte (BGE) with negative-charged mixed micelles as additive. The method was successively applied to urine samples. The LOD and LOQ for Tobramycin in urine were 90 and 200 ng/ml respectively and recovery was >98% (n = 5). All urine samples were analyzed by direct injection without sample pre-treatment. Another use of hyphenated analytical technique, capillary zone electrophoresis (CZE) connected to ultraviolet (UV) detector was also used for sensitive analysis of rhEPO at low levels (2000 IU) in the presence of large amount of human serum albumin (HSA). Analysis of rhEPO was achieved by the use of the electrokinetic injection (EI) with discontinuous buffers. Phosphate buffer was used as BGE with metal ions as additive. The proposed method can be used for the estimation of large number of quality control rhEPO samples in a short period.

  8. Study of the gamma spectrum of 16N with a BGO detector, for the purpose of calibration and of determining the fluorine grade of mineral samples

    International Nuclear Information System (INIS)

    Castro-Garcia, M.P.; Alonso-Sanchez, T.; Rey-Ronco, M.A.

    2013-01-01

    The study of 16 N's gamma spectrum has two main uses: calibrating gamma detectors in a high energy range, and determining the fluorine grade of mineral samples of fluorite. This article examines and compares the gamma ray spectrum of 16 N as recorded by a Bi 4 GeO 12 detector, as well as the resolution of this detector at high energy levels, and the signal-background relationship of an experimental laboratory cyclic activation unit. 16 N is the product of the reaction 9-F-19(n,α)7-N-16, which takes place during the neutron activation of mineral samples of fluorspar, and its production depends, among other factors, upon the grade of fluorite. The technique used in this study is cyclic-type neutron activation for recording delayed gamma rays, carried out with an americium-beryllium neutron source with an activity of 1 Ci. Lastly, a correlation is established between the area below the peak amount of 16 N emitted by the sample, and the sample's fluorite grade. (author)

  9. A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

    Energy Technology Data Exchange (ETDEWEB)

    C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

    1998-10-01

    The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

  10. Radiation shielding activities at IDOM

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  11. Radiation shielding activities at IDOM

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  12. Analytic flux formulas and tables of shielding functions

    International Nuclear Information System (INIS)

    Wallace, O.J.

    1981-06-01

    Hand calculations of radiation flux and dose rates are often useful in evaluating radiation shielding and in determining the scope of a problem. The flux formulas appropriate to such calculations are almost always based on the point kernel and allow for at most the consideration of laminar slab shields. These formulas often require access to tables of values of integral functions for effective use. Flux formulas and function tables appropriate to calculations involving homogeneous source regions with the shapes of lines, disks, slabs, truncated cones, cylinders, and spheres are presented. Slab shields may be included in most of these calculations, and the effect of a cylindrical shield surrounding a cylindrical source may be estimated. Detector points may be located axially, laterally, or interior to a cylindrical source. Line sources may be tilted with respect to a slab shield. All function tables are given for a wide range of arguments

  13. Combination thermal and radiation shield for well logging apparatus

    International Nuclear Information System (INIS)

    Wilson, B.F.

    1984-01-01

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

  14. Investigations of some building materials for γ-rays shielding effectiveness

    Science.gov (United States)

    Mann, Kulwinder Singh; Kaur, Baljit; Sidhu, Gurdeep Singh; Kumar, Ajay

    2013-06-01

    For construction of residential and non-residential buildings bricks are used as building blocks. Bricks are made from mixtures of sand, clay, cement, fly ash, gypsum, red mud and lime. Shielding effectiveness of five soil samples and two fly ash samples have been investigated using some energy absorption parameters (Mass attenuation coefficients, mass energy absorption coefficients, KERMA (kinetic energy released per unit mass), HVL, equivalent atomic number and electron densities) firstly at 14 different energies from 81-1332 keV then extended to wide energy range 0.015-15 MeV. The soil sample with maximum shielding effectiveness has been used for making eight fly ash bricks [(Lime)0.15 (Gypsum)0.05 (Fly Ash)x (Soil)0.8-x, where values of x are from 0.4-0.7]. High Purity Germanium (HPGe) detector has been used for gamma-ray spectroscopy. The elemental compositions of samples were analysed using an energy dispersive X-ray fluorescence (EDXRF) spectrometer. The agreements of theoretical and experimental values of mass attenuation coefficient have been found to be quite satisfactory. It has been verified that common brick possess the maximum shielding effectiveness for wide energy range 0.015-15 MeV. The results have been shown graphically with some useful conclusions for making radiation safe buildings.

  15. Neutron detector for detecting rare events of spontaneous fission

    International Nuclear Information System (INIS)

    Ter-Akop'yan, G.M.; Popeko, A.G.; Sokol, E.A.; Chelnokov, L.P.; Smirnov, V.I.; Gorshkov, V.A.

    1981-01-01

    The neutron detector for registering rare events of spontaneous fission by detecting multiple neutron emission is described. The detector represents a block of plexiglas of 550 mm diameter and 700 mm height in the centre of which there is a through 160 mm diameter channel for the sample under investigation. The detector comprises 56 3 He filled counters (up to 7 atm pressure) with 1% CO 2 addition. The counters have a 500 mm length and a 32 mm diameter. The sampling of fission events is realized by an electron system which allows determining the number of detected neutrons, numbers of operated counters, signal amplitude and time for fission event detecting. A block diagram of a neutron detector electron system is presented and its operation principle is considered. For protection against cosmic radiation the detector is surronded by a system of plastic scintillators and placed behind the concrete shield of 6 m thickness. The results of measurements of background radiation are given. It has been found that the background radiation of single neutron constitutes about 150 counts per hour, the detecting efficiency of single neutron equals 0.483 +- 0.005, for a 10l detector sensitive volume. By means of the detector described the parameters of multiplicity distribution of prompt neutrons for 256 Fm spontaneous fission are measured. The average multiplicity equals 3.59+-0.06 the dispersion being 2.30+-0.65

  16. Shielding member for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori

    1997-06-30

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

  17. Study of gamma irradiation effects on the etching and optical properties of CR-39 solid state nuclear track detector and its application to uranium assay in soil samples

    International Nuclear Information System (INIS)

    Amol Mhatre; Kalsi, P.C.

    2011-01-01

    The gamma irradiation effects in the dose range of 2.5-43.0 Mrad on the etching and optical characteristics of CR-39 solid state nuclear track detector (SSNTD) have been studied by using etching and UV-Visible spectroscopic techniques. From the measured bulk etch rates at different temperatures, the activation energies for bulk etching at different doses have also been determined. It is seen that the bulk etch rates increase and the activation energies for bulk etching decrease with the increase in gamma dose. The optical band gaps of the unirradiated and the gamma -irradiated detectors determined from the UV-Visible spectra were found to decrease with the increase in gamma dose. These results have been explained on the basis of scission of the detector due to gamma irradiation. The present studies can be used for the estimation of gamma dose in the range of 2.5-43.0 Mrad and can also be used for estimating track registration efficiency in the presence of gamma dose. The CR-39 detector has also been applied for the assay of uranium in some soil samples of Jammu city. (author)

  18. Estimating ISABELLE shielding requirements

    International Nuclear Information System (INIS)

    Stevens, A.J.; Thorndike, A.M.

    1976-01-01

    Estimates were made of the shielding thicknesses required at various points around the ISABELLE ring. Both hadron and muon requirements are considered. Radiation levels at the outside of the shield and at the BNL site boundary are kept at or below 1000 mrem per year and 5 mrem/year respectively. Muon requirements are based on the Wang formula for pion spectra, and the hadron requirements on the hadron cascade program CYLKAZ of Ranft. A muon shield thickness of 77 meters of sand is indicated outside the ring in one area, and hadron shields equivalent to from 2.7 to 5.6 meters in thickness of sand above the ring. The suggested safety allowance would increase these values to 86 meters and 4.0 to 7.2 meters respectively. There are many uncertainties in such estimates, but these last figures are considered to be rather conservative

  19. Technological advances in cosmogenic neutron detectors for measuring soil water content

    Science.gov (United States)

    Zreda, M. G.; Schrön, M.; Köhli, M.

    2017-12-01

    The cosmic-ray neutron probe is used for measuring area-average soil water content at the hectometer scale. Early work showed a simple exponential decrease with distance of the instrument's sensitivity and a footprint 300 m in radius. Recent research suggested a much higher sensitivity to local neutrons and reduced footprint. We show results confirming the high sensitivity to local neutrons, describe two ways to reduce local and increase far-field effects, and propose ways of measuring neutrons at different spatial scales. Measurements with moderated detectors across a 10-m-wide creek and a 2-m-wide water tank show a decrease by 30% and 20%, respectively, of neutron intensity over water compared to that over land nearby. These results mean that the detector is sensitive to meter-scale heterogeneities of water content. This sensitivity can be reduced by rising the detector or by shielding it from local neutrons. The effect of local water distributions on the measured neutron intensity decreases with height. In the water tank experiment it disappeared almost completely at the height of 2 m, leading to the conjecture that the height roughly equal to the horizontal scale of heterogeneity would eliminate the sensitivity. This may or may not be practical. Shielding the detector below by a hydrogenous material removes a substantial fraction of the local neutrons. The shielded detector has a reduced count rate, reduced sensitivity to local neutrons and increased sensitivity to neutrons farther afield, and a larger footprint. Such a detector could be preferable to the current cosmogenic-neutron probe under heterogeneous soil water conditions. The shielding experiments also inspired the development of a local-area neutron detector. It has hydrogenous neutron shields on all sides except the bottom, substantially blocking the neutrons coming from afar, while allowing the neutrons coming directly from below. Its footprint is equal to its physical dimension when the detector is

  20. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

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

  1. Radiation shielding curtain

    International Nuclear Information System (INIS)

    Winkler, N.T.

    1976-01-01

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

  2. Shielded cells transfer automation

    International Nuclear Information System (INIS)

    Fisher, J.J.

    1984-01-01

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

  3. Electromagnetic shielding effectiveness of 3D printed polymer composites

    Science.gov (United States)

    Viskadourakis, Z.; Vasilopoulos, K. C.; Economou, E. N.; Soukoulis, C. M.; Kenanakis, G.

    2017-12-01

    We report on preliminary results regarding the electromagnetic shielding effectiveness of various 3D printed polymeric composite structures. All studied samples were fabricated using 3D printing technology, following the fused deposition modeling approach, using commercially available filaments as starting materials. The electromagnetic shielding performance of the fabricated 3D samples was investigated in the so called C-band of the electromagnetic spectrum (3.5-7.0 GHz), which is typically used for long-distance radio telecommunications. We provide evidence that 3D printing technology can be effectively utilized to prepare operational shields, making them promising candidates for electromagnetic shielding applications for electronic devices.

  4. High resolution imaging of 2D distribution of lithium in thin samples measured with multipixel detectors in sandwich geometry

    Czech Academy of Sciences Publication Activity Database

    Tomandl, Ivo; Vacík, Jiří; Sierra, Y. M.; Granja, C.; Kraus, V.

    2017-01-01

    Roč. 88, č. 2 (2017), č. článku 023706. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : Li * detectors * thermal neutrons Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 1.515, year: 2016

  5. Use of semiconductor detector c-Si microstrip type in obtaining the digital radiographic imaging of phantoms and biological samples of mammary glands

    International Nuclear Information System (INIS)

    Leyva, A.; Cabal, A.; Pinera, I.; Abreu, Y.; Cruz, C. M.; Montano, L. M.; Diaz, C. C.; Fontaine, M.; Ortiz, C. M.; Padilla, F.; De la Mora, R.

    2009-01-01

    The present work synthesizes the experimental results obtained in the characterization of 64 micro strips crystalline silicon detector designed for experiments in high energies physics, with the objective of studying its possible application in advanced medical radiography, specifically in digital mammography and angiography. The research includes the acquisition of two-dimensional radiography of a mammography phantom using the scanning method, and its comparison with similar images simulated mathematically for different X rays sources. The paper also shows the experimental radiography of two biological samples taken from biopsies of mammas, where it is possible to identify the presence of possible pathological lesions. The results reached in this work point positively toward the effective possibility of satisfactorily introducing those advanced detectors in medical digital imaging applications. (Author)

  6. Mobility and powering of large detectors. Moving large detectors

    International Nuclear Information System (INIS)

    Thompson, J.

    1977-01-01

    The possibility is considered of moving large lepton detectors at ISABELLE for readying new experiments, detector modifications, and detector repair. A large annex (approximately 25 m x 25 m) would be built adjacent to the Lepton Hall separated from the Lepton Hall by a wall of concrete 11 m high x 12 m wide (for clearance of the detector) and approximately 3 m thick (for radiation shielding). A large pad would support the detector, the door, the cryogenic support system and the counting house. In removing the detector from the beam hall, one would push the pad into the annex, add a dummy beam pipe, bake out the beam pipe, and restack and position the wall on a small pad at the door. The beam could then operate again while experimenters could work on the large detector in the annex. A consideration and rough price estimate of various questions and proposed solutions are given

  7. Application of a dummy eye shield for electron treatment planning

    International Nuclear Information System (INIS)

    Kang, Sei-Kwon; Park, Soah; Hwang, Taejin; Cheong, Kwang-Ho; Han, Taejin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Oh, Do Hoon; Bae, Hoonsik

    2013-01-01

    Metallic eye shields have been widely used for near-eye treatments to protect critical regions, but have never been incorporated into treatment plans because of the unwanted appearance of the metal artifacts on CT images. The purpose of this work was to test the use of an acrylic dummy eye shield as a substitute for a metallic eye shield during CT scans. An acrylic dummy shield of the same size as the tungsten eye shield was machined and CT scanned. The BEAMnrc and the DOSXYZnrc were used for the Monte Carlo (MC) simulation, with the appropriate material information and density for the aluminum cover, steel knob and tungsten body of the eye shield. The Pinnacle adopting the Hogstrom electron pencil-beam algorithm was used for the one-port 6-MeV beam plan after delineation and density override of the metallic parts. The results were confirmed with the metal oxide semiconductor field effect transistor (MOSFET) detectors and the Gafchromic EBT2 film measurements. For both the maximum eyelid dose over the shield and the maximum dose under the shield, the MC results agreed with the EBT2 measurements within 1.7%. For the Pinnacle plan, the maximum dose under the shield agreed with the MC within 0.3%; however, the eyelid dose differed by -19.3%. The adoption of the acrylic dummy eye shield was successful for the treatment plan. However, the Pinnacle pencil-beam algorithm was not sufficient to predict the eyelid dose on the tungsten shield, and more accurate algorithms like MC should be considered for a treatment plan. (author)

  8. Design of a software for gamma detector efficiency

    International Nuclear Information System (INIS)

    Lopez, G.

    2011-01-01

    Gamma spectroscopy with highly-pure-germanium detector is one of the most used method for qualitative and quantitative analysis of samples. Nevertheless Gamma spectroscopy results require to be corrected, first for taking into account the self-shielding effect that represents the absorption of the photons by the sample itself and secondly for correcting the fact that 2 photons emitted simultaneously with energy E 1 and E 2 are likely to be simultaneously detected and then counted as a single photon with an energy E 1 +E 2 . This effect is called gamma-gamma coincidence. A software has been designed to simulate both effect and produce correcting factors in the case of cylindrical geometries. This software has been validated on Americium 241 for the self-shielding effect and on Cesium 134 for gamma-gamma coincidence. (A.C.)

  9. Applying a low energy HPGe detector gamma ray spectrometric technique for the evaluation of Pu/Am ratio in biological samples.

    Science.gov (United States)

    Singh, I S; Mishra, Lokpati; Yadav, J R; Nadar, M Y; Rao, D D; Pradeepkumar, K S

    2015-10-01

    The estimation of Pu/(241)Am ratio in the biological samples is an important input for the assessment of internal dose received by the workers. The radiochemical separation of Pu isotopes and (241)Am in a sample followed by alpha spectrometry is a widely used technique for the determination of Pu/(241)Am ratio. However, this method is time consuming and many times quick estimation is required. In this work, Pu/(241)Am ratio in the biological sample was estimated with HPGe detector based measurements using gamma/X-rays emitted by these radionuclides. These results were compared with those obtained from alpha spectroscopy of sample after radiochemical analysis and found to be in good agreement. Copyright © 2015 Elsevier Ltd. All rights reserved.

  10. Glasses impregnated with lead for radiation shielding

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  11. Attenuation of fast neutron in concretes for biological shielding

    International Nuclear Information System (INIS)

    Labrada, A.; Chavez, A.; Gonzalez Mateu, D.; Desdin, F.; Tenjeiro, J.I.; Tellez, E.

    1993-01-01

    The attenuation of neutrons emitted by an 10 6 n/s. Am-Be source, in concretes elaborated with different aggregates is discussed in this paper. Two measurement methods were used an dosimetric system with Bonner spheres and 6 LiI(Eu) detector, and LAVSAN dielectric nuclear track detectors - with 238 U converts. The concretes elaborated with magnetite is reported as the best for neutron shielding while the Bauxite is not advisable for this purpose

  12. Neutron detector assembly

    International Nuclear Information System (INIS)

    Hanai, Koi; Shirayama, Shinpei.

    1978-01-01

    Purpose: To prevent gamma-ray from leaking externally passing through the inside of a neutron detector assembly. Constitution: In a neutron detector assembly having a protection pipe formed with an enlarged diameter portion which serves also as a spacer, partition plates with predetermined width are disposed at the upper and the lower portions in this expanded portion. A lot of metal particles are filled into spaces formed by the partition plates. In such a structure, the metal particles well-absorb the gamma-rays from above and convert them into heat to provide shielding for the gamma-rays. (Horiuchi, T.)

  13. Neutron shielding material

    International Nuclear Information System (INIS)

    Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

    1993-01-01

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  14. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

    The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

  15. Method for dismantling shields

    International Nuclear Information System (INIS)

    Fukuzawa, Rokuro; Kondo, Nobuhiro; Kamiyama, Yoshinori; Kawasato, Ken; Hiraga, Tomoaki.

    1990-01-01

    The object of the present invention is to enable operators to dismantle shieldings contaminated by radioactivity easily and in a short period of time without danger of radiation exposure. A plurality of introduction pipes are embedded previously to the shielding walls of shielding members which contain a reactor core in a state where both ends of the introduction pipes are in communication with the outside. A wire saw is inserted into the introduction pipes to cut the shieldings upon dismantling. Then, shieldings can be dismantled easily in a short period of time with no radiation exposure to operator's. Further, according to the present invention, since the wire saw can be set easily and a large area can be cut at once, operation efficiency is improved. Further, since remote control is possible, cutting can be conducted in water and complicated places of the reactor. Biting upon starting the wire saw in the introduction pipe is reduced to facilitate startup for the rotation. (I.S.)

  16. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Female gonadal shielding with automatic exposure control increases radiation risks

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-02-15

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

  18. Female gonadal shielding with automatic exposure control increases radiation risks

    International Nuclear Information System (INIS)

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

    2018-01-01

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

  19. Female gonadal shielding with automatic exposure control increases radiation risks.

    Science.gov (United States)

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

    2018-02-01

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

  20. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  1. Shield support frame. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Plaga, K.

    1981-09-17

    A powered shield support frame for coal sheds is described comprising of two bottom sliding shoes, a large area gob shield and a larg area roof assembly, all joined movable together. The sliding shoes and the gob shield are joined by a lemniscate guide. Two hydraulic props are arranged at the face-side at one third of the length of the sliding shoes and at the goaf-side at one third of the length of the roof assembly. A nearly horizontal lying pushing prop unit joins the bottom wall sliding shoes to the goaf-side lemniscate guide. This assembly can be applied to seams with a thickness down to 45 cm. (OGR).

  2. Radiation shielding material

    International Nuclear Information System (INIS)

    Matsumoto, Akio; Isobe, Eiji.

    1976-01-01

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

  3. Hybrid Magnetic Shielding

    Science.gov (United States)

    Royal, Kevin; Crawford, Christopher; Mullins, Andrew; Porter, Greg; Blanton, Hunter; Johnstone, Connor; Kistler, Ben; Olivera, Daniela

    2017-09-01

    The search for the electric dipole moment of the neutron requires the ambient magnetic field to be on the pT scale which is accomplished with large magnetic shielding rooms. These rooms are fitted with large mu-metal sheets to allow for passive cancellation of background magnetic fields. Active shielding technology cannot uniformly cancel background magnetic fields. These issues can be remedied by combining the methods into a hybrid system. The design used is composed of panels that have an active layer of cancellation between two sheets of mu-metal. The panels form a cube and draw in magnetic fields perpendicular to the surface which can then be reduced using active shielding. This work is supported by the Department of Energy under Contract DE-SC0008107.

  4. Reactor head shielding apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

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

  6. Radiation detectors

    International Nuclear Information System (INIS)

    2013-01-01

    This sixth chapter presents the operational principles of the radiation detectors; detection using photographic emulsions; thermoluminescent detectors; gas detectors; scintillation detectors; liquid scintillation detectors; detectors using semiconductor materials; calibration of detectors; Bragg-Gray theory; measurement chain and uncertainties associated to measurements

  7. Testing and Performance Validation of a Shielded Waste Segregation and Clearance Monitor Designed for the Measurement of Low Level Waste-13043

    International Nuclear Information System (INIS)

    Mason, John A.; Burke, Kevin J.; Towner, Antony C.N.; Beaven, Graham; Spence, Robert

    2013-01-01

    This paper describes the development, testing and validation of a shielded waste segregation and clearance monitor designed for the measurement of low-density low-level waste (LLW). The monitor is made of a measurement chamber surrounded by detectors and a shielded outer frame. The shielded chamber consists of a steel frame, which contains typically 1.5 inches (3.81 cm) of lead and 0.5 inches (1.27 cm) of steel shielding. Inside the shielding are plastic scintillator panels, which serve as gross gamma ray detectors. The detector panels, with embedded photomultipliers, completely surround the internal measurement chamber on all 6 sides. Care has been taken to distribute the plastic scintillator detectors in order to optimise both the efficiency for gamma ray detection and at the same time achieve a volumetric sensitivity, which is as uniform as possible. A common high voltage power supply provides the bias voltage for each of the six photomultipliers. The voltage signals arising from the detectors and photomultipliers are amplified by six sensitive amplifiers. Each amplifier incorporates a single channel analyser with both upper and lower thresholds and the digitised counts from each detector are recorded on six scalars. Operation of the device is by means of a microprocessor from which the scalars are controlled. An internal load cell linked to the microprocessor determines the weight of the waste object, and this information is used to calculate the specific activity of the waste. The monitor makes background measurements when the shielded door is closed and a sample, usually a bag of low-density waste, is not present in the measurement chamber. Measurements of the minimum detectable activity (MDA) of an earlier large volume prototype instrument are reported as part of the development of the Waste Segregation and Clearance Monitor (WSCM) described in the paper. For the optimised WSCM a detection efficiency of greater than 32% was measured using a small Cs-137

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Radiation shielding bricks

    International Nuclear Information System (INIS)

    Crowe, G.J.W.

    1983-01-01

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

  10. Flux trapping and shielding in irreversible superconductors

    International Nuclear Information System (INIS)

    Frankel, D.J.

    1978-05-01

    Flux trappings and shielding experiments were carried out on Pb, Nb, Pb-Bi, Nb-Sn, and Nb-Ti samples of various shapes. Movable Hall probes were used to measure fields near or inside the samples as a function of position and of applied field. The trapping of transverse multipole magnetic fields in tubular samples was accomplished by cooling the samples in an applied field and then smoothly reducing the applied field to zero. Transverse quadrupole and sextupole fields with gradients of over 2000 G/cm were trapped with typical fidelity to the original impressed field of a few percent. Transverse dipole fields of up to 17 kG were also trapped with similar fidelity. Shielding experiments were carried out by cooling the samples in zero field and then gradually applying an external field. Flux trapping and shielding abilities were found to be limited by two factors, the pinning strength of the material, and the susceptibility of a sample to flux jumping. The trapping and shielding behavior of flat disk samples in axial fields and thin-walled tubular samples in transverse fields was modeled. The models, which were based on the concept of the critical state, allowed a connection to be made between the pinning strength and critical current level, and the flux trapping and shielding abilities. Adiabatic and dynamic stability theories are discussed and applied to the materials tested. Good qualitative, but limited quantitative agreement was obtained between the predictions of the theoretical stability criteria and the observed flux jumping behavior

  11. Hybrid Active-Passive Radiation Shielding System

    Data.gov (United States)

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

  12. Sampling

    CERN Document Server

    Thompson, Steven K

    2012-01-01

    Praise for the Second Edition "This book has never had a competitor. It is the only book that takes a broad approach to sampling . . . any good personal statistics library should include a copy of this book." —Technometrics "Well-written . . . an excellent book on an important subject. Highly recommended." —Choice "An ideal reference for scientific researchers and other professionals who use sampling." —Zentralblatt Math Features new developments in the field combined with all aspects of obtaining, interpreting, and using sample data Sampling provides an up-to-date treat

  13. Energy dispersive X-ray fluorescence from useless tyres samples with a Si PIN detector; Fluorescencia de raios X por dispersao em energia de amostras de pneus inserviveis com detector de Si-PIN

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Fabio; Scheibel, Viviane [Universidade Estadual de Londrina, PR (Brazil). Dept. de Fisica. Lab. de Fisica Nuclear Aplicada]. E-mail: bonn@uel.br; Melquiades, Fabio Luiz [Universidade Estadual do Centro-Oeste, Guarapuava, PR (Brazil). Dept. de Fisica; Moraes, Liz Mary Bueno de [Centro de Energia Nuclear na Agricultura (CENA), Piracicaba, SP (Brazil). Lab. de Instrumentacao Nuclear

    2005-07-01

    The concentration of Zn from discard tyre samples is of environmental interest, since on its production are used S for the rubber vulcanization process, and Zn O as reaction catalyze. The useless tyres are been used for asphalt pave, burn in cement industry and thermoelectric power plant and in erosion control of agriculture areas. Analyses of these samples requires frequently chemical digestion that is expensive and take a long time. Trying to eliminate these limitations, the objective of this work was use Energy Dispersive X Ray Fluorescence technique (EDXRF) with a portable system as the technique is multi elementary and needs a minimum sample preparation. Five useless tyres samples were grind in a knife mill and after this in a cryogenic mill, and analyzed in pellets form, using a X ray mini tube (Ag target, Mo {sub l}ter, 25 kV/20 {sub A}) for 200 s and a Si-PIN semiconductor detector coupled to a multichannel analyzer. Were obtained Zn concentrations in the range of 40.6 to 44.2 {sub g} g{sub 1}, representing nearly 0.4. (author)

  14. Energy-dispersive X-ray fluorescence of discarded tire samples, using a Si-PIN detector; Fluorescencia de raios X por dispersao em energia de amostras de pneus inserviveis com detector de Si-Pin

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Fabio; Appoloni, C.R., E-mail: bonn@uel.b [Universidade Estadual de Londrina (UEL), PR (Brazil). Dept. de Fisica. Lab. de Fisica Nuclear Aplicada; Melquiades, Fabio L. [Universidade Estadual do Centro Oeste (UNICENTRO), Guarapuava, PR (Brazil). Dept. de Fisica

    2007-07-01

    The determination of zinc concentration in samples of discarded tires is of great environmental interest because the process for manufacturing tyres uses S for rubber vulcanization, and ZnO is the reaction catalyst. Discarded tyres are being used in asphalt paving, in the burning process of thermoelectric and cement industries and also for controlling erosion in agricultural areas. Analysis of tyre samples usually requires chemical digestion which is slow and expensive. Aiming to eliminate those limitations, this work uses energy-dispersive X-ray fluorescence (EDXRF) with a portable equipment, once it is a simultaneous multi-element analytical technique, requiring minimal sample preparation. Five samples of discarded tyres have been ground and analysed in the form of pastilles, using a mini X-ray tube (Ag target, MO filter, 25 kV/20 muA) for 200 s, and a Si-PIN semiconductor detector coupled to a multichannel analyser. Zinc concentrations in the range of 40.6 to 44.2 mug g{sup -1} have been obtained, representing 0.4% of the tire composition, which is below the maximum value (2%) recommended by the European Tyre Recycling Association. Concentrations between 0.15 and 0.52 mug g{sup -1} were obtained for Fe

  15. Radiation shielding cloth

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Electrostatic shielding of transformers

    Energy Technology Data Exchange (ETDEWEB)

    De Leon, Francisco

    2017-11-28

    Toroidal transformers are currently used only in low-voltage applications. There is no published experience for toroidal transformer design at distribution-level voltages. Toroidal transformers are provided with electrostatic shielding to make possible high voltage applications and withstand the impulse test.

  17. Penetration portion shielding structure

    International Nuclear Information System (INIS)

    Hayashi, Katsumi; Narita, Hitoshi; Handa, Hiroyuki; Takeuchi, Jun; Tozuka, Fumio.

    1994-01-01

    Openings of a plurality of shieldings for penetration members are aligned to each other, and penetration members are inserted from the openings. Then, the openings of the plurality of shielding members are slightly displaced with each other to make the penetration portions into a helical configuration, so that leakage of radiation is reduced. Upon removal of the members, reverse operation is conducted. When a flowable shielding material is used, the penetration portions are constituted with two plates having previously formed openings and pipes for connecting the openings with each other and a vessel covering the entire of them. After passing the penetration members such as a cable, the relative position of the two plates is changed by twisting, to form a helical configuration which reduces radiation leakage. Since they are bent into the helical configuration, shielding performance is extremely improved compared with a case that radiation leakage is caused from an opening of a straight pipe. In addition, since they can be returned to straight pipes, attachment, detachment and maintenance can be conducted easily. (N.H.)

  18. Radiation shielding glass

    International Nuclear Information System (INIS)

    Kido, Kazuhiro; Ueda, Hajime.

    1997-01-01

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

  19. Neutron streaming studies along JET shielding penetrations

    Science.gov (United States)

    Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

    2017-09-01

    Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

  20. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

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

    1980-01-01

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  1. Neutron multicounter detector for investigation of content and spatial distribution of fission materials in large volume samples

    International Nuclear Information System (INIS)

    Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

    1998-01-01

    The experimental device is a neutron coincidence well counter. It can be applied for passive assay of fissile - especially for plutonium bearing - materials. It consist of a set of 3 He tubes placed inside a polyethylene moderator; outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using neutron correlator connected with a PC, and correlation techniques implemented in software. Such a neutron counter allows for determination of plutonium mass ( 240 Pu effective mass) in nonmultiplying samples having fairly big volume (up to 0.14 m 3 ). For determination of neutron sources distribution inside the sample, the heuristic methods based on hierarchical cluster analysis are applied. As an input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples, are taken. Such matrices are collected by means of sample scanning by detection head. During clustering process, counts profiles for unknown samples fitted into dendrograms using the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in an examined sample is then evaluated on the basis of comparison with standard sources distributions. (author)

  2. Measurement of the $M_A^{QE}$ parameter using multiple quasi-elastic dominated sub-samples in the minos near detector

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Nathan Samuel [Indiana Univ., Bloomington, IN (United States)

    2011-12-05

    The Main Injector Neutrino Oscillation Search (MINOS) is a two detector, long baseline neutrino oscillation experiment. The MINOS near detector is an ironscintillator tracking/sampling calorimeter and has recorded the world’s largest data set of neutrino interactions in the 0-5 GeV region. This high statistics data set is used to make precision measurements of neutrino interaction cross-sections on iron. The Q2 dependence in charged current quasi-elastic (CCQE) scattering probes the axial and vector structure (form factor) of the nucleon/nuclear target, and nuclear effects in neutrino scattering. Presented here is a study of the MINOS Data that will introduce a method that improves the existing MINOS CCQE analysis. This analysis uses an additional CCQE dominated sub-sample from a different kinematic region to reduce correlations between fit parameters in the existing MINOS CCQE analysis. The measured value of the axial-vector mass is MQE A = 1.312+0.037 -0.038(fit)+0.123 -0.265(syst.) GeV.

  3. Shielding Effectiveness Measurements using a Reverberation Chamber

    NARCIS (Netherlands)

    Leferink, Frank Bernardus Johannes; Bergsma, J.G.; Bergsma, Hans; van Etten, Wim

    2006-01-01

    Shielding effectiveness measurements have been performed using a reverberation chamber. The reverberation chamber methodology as we1l as the measurement setup is described and some results are given. Samples include glass reinforced plastic panels, aluminum panels with many holes, wire mesh, among

  4. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions

  5. Shielding experiments for accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  6. Shielding experiments for accelerator facilities

    International Nuclear Information System (INIS)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

    2000-01-01

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

  7. Development of shielding design analysis system

    International Nuclear Information System (INIS)

    Tada, Keiko; Shiraki, Takako

    2001-03-01

    The aim of this work is to develop insufficient auxiliary routines which manage input and output data and interface the main codes and to establish a shielding design analysis system on work stations (SUN, DEC). In shielding design analyses, one- and two- dimensional (1-D and 2-D) transport Sn codes are used mainly with some auxiliary codes which generate input data of Sn calculation and edit Sn calculation outputs. The main transport calculation codes can be obtained from the Code Center of RIST (Research Organization for Information Science and Technology). In this work, peripheral codes are developed to generate cross sections, produce Sn quadrature sets, edit calculation outputs or draw contour figures. In shielding calculations around a reactor, the boot-strapping technique is often employed to treat a large area extending from the core to the biological shield to improve the calculation accuracy. When a three-dimensional (3-D) calculation for a complex geometry with shielding defects, 2-D and 3-D coupling calculation is employed frequently. To use this coupling method conversion cods are prepared which read flux file from DORT and prepare an external boundary source file for the 2-D or the 3-D calculation codes. For further conveniences well used data such as the Sn quadrature sets, the dose rate conversion factors, the reaction cross section sets are stored as a data base and code manuals including sample inputs of typical problems are prepared which are comprehensible to beginners. (author)

  8. Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

    International Nuclear Information System (INIS)

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

    2015-01-01

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

  9. Effects of increased shielding on gamma-radiation levels within spacecraft

    Science.gov (United States)

    Haskins, P. S.; McKisson, J. E.; Weisenberger, A. G.; Ely, D. W.; Ballard, T. A.; Dyer, C. S.; Truscott, P. R.; Piercey, R. B.; Ramayya, A. V.; Camp, D. C.

    The Shuttle Activation Monitor (SAM) experiment was flown on the Space Shuttle Columbia (STS-28) from 8 - 13 August, 1989 in a 57°, 300 km orbit. One objective of the SAM experiment was to determine the relative effect of different amounts of shielding on the gamma-ray backgrounds measured with similarly configured sodium iodide (NaI) and bismuth germante (BGO) detectors. To achieve this objective twenty-four hours of data were taken with each detector in the middeck of the Shuttle on the ceiling of the airlock (a high-shielding location) as well as on the sleep station wall (a low-shielding location). For the cosmic-ray induced background the results indicate an increased overall count rate in the 0.2 to 10 MeV energy range at the more highly shielded location, while in regions of trapped radiation the low shielding configuration gives higher rates at the low energy end of the spectrum.

  10. Self powered platinum flux detector application for shutdown system

    International Nuclear Information System (INIS)

    Su Guoquan

    2005-01-01

    This article introduce Neutron Flux Detector application in Candu Power Plant, including: design purpose, location in the site, dynamic compensation, differential compensation, detector assembly pressurized with high pure helium etc. And shielding grounding improvement is suggested because of detector signal and setpoint signal noise. (authors)

  11. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

    A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

  12. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  13. Shielding benchmark test

    International Nuclear Information System (INIS)

    Kawai, Masayoshi

    1984-01-01

    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)

  14. Neutron shielding material

    International Nuclear Information System (INIS)

    Suzuki, Shigenori; Iimori, Hiroshi; Kobori, Junzo.

    1980-01-01

    Purpose: To provide a neutron shielding material which incorporates preferable shielding capacity, heat resistance, fire resistance and workability by employing a mixture of thermosetting resin, polyethylene and aluminium hydroxide in special range ratio and curing it. Constitution: A mixture containing 20 to 60% by weight of thermosetting resin having preferable heat resistance, 10 to 40% by weight of polyethylene powder having high hydrogen atom density and 1000 to 60000 of molecular weight, and 15 to 55% by weight of Al(OH) 3 for imparting fire resistance and self-fire extinguishing property thereto is cured. At this time approx. 0.5 to 5% of curing catalyst of the thermosetting resin is contained in 100 parts by weight of the mixture. (Sekiya, K.)

  15. Radiation shielding wall structure

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. 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.

  17. Light shielding apparatus

    Science.gov (United States)

    Miller, Richard Dean; Thom, Robert Anthony

    2017-10-10

    A light shielding apparatus for blocking light from reaching an electronic device, the light shielding apparatus including left and right support assemblies, a cross member, and an opaque shroud. The support assemblies each include primary support structure, a mounting element for removably connecting the apparatus to the electronic device, and a support member depending from the primary support structure for retaining the apparatus in an upright orientation. The cross member couples the left and right support assemblies together and spaces them apart according to the size and shape of the electronic device. The shroud may be removably and adjustably connectable to the left and right support assemblies and configured to take a cylindrical dome shape so as to form a central space covered from above. The opaque shroud prevents light from entering the central space and contacting sensitive elements of the electronic device.

  18. Shielding container for radioactive isotopes

    International Nuclear Information System (INIS)

    Sumi, Tetsuo; Tosa, Masayoshi; Hatogai, Tatsuaki.

    1975-01-01

    Object: To effect opening and closing bidirectional radiation used particularly for a gamma densimeter or the like by one operation. Structure: This device comprises a rotatable shielding body for receiving radioactive isotope in the central portion thereof and having at least two radiation openings through which radiation is taken out of the isotope, and a shielding container having openings corresponding to the first mentioned radiation openings, respectively. The radioactive isotope is secured to a rotational shaft of the shielding body, and the shielding body is rotated to register the openings of the shielding container with the openings of the shielding body or to shield the openings, thereby effecting radiation and cut off of gamma ray in the bidirection by one operation. (Kamimura, M.)

  19. Primary shield displacement and bowing

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    The reactor primary shield is constructed of high density concrete and surrounds the reactor core. The inlet, outlet and side primary shields were constructed in-place using 2.54 cm (1 in) thick steel plates as the forms. The plates remained as an integral part of the shields. The elongation of the pressure tubes due to thermal expansion and pressurization is not moving through the inlet nozzle hardware as designed but is accommodated by outward displacement and bowing of the inlet and outlet shields. Excessive distortion of the shields may result in gas seal failures, intolerable helium gas leaks, increased argon-41 emissions, and shield cooling tube failures. The shield surveillance and testing results are presented

  20. Shielding calculations using FLUKA

    International Nuclear Information System (INIS)

    Yamaguchi, Chiri; Tesch, K.; Dinter, H.

    1988-06-01

    The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)

  1. Shielding Benchmark Computational Analysis

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. Muon shielding for PEP

    International Nuclear Information System (INIS)

    Jenkins, T.M.; Thomas, R.H.

    1974-01-01

    The first stage of construction of PEP will consist of electron and positron storage rings. At a later date a 200 GeV proton storage ring may be added. It is judicious therefore, to ensure that the first and second phases of construction are compatible with each other. One of several factors determining the elevation at which the storage rings will be constructed is the necessity to provide adequate radiation shielding. The overhead shielding of PEP is determined by the reproduction of neutrons in the hadron cascade generated by primary protons lost from the storage ring. The minimum overburden planned for PEP is 5.5 meters of earth (1100 gm cm/sup /minus/2/). To obtain a rough estimate of the magnitude of the muon radiation problem this note presents some preliminary calculations. Their purpose is intended merely to show that the presently proposed design for PEP will present no major shielding problems should the protons storage ring be installed. More detailed calculations will be made using muon yield computer codes developed at CERN and NAL and muon transport codes developed at SLAC, when details of the proton storage ring become settled. 9 refs., 4 figs

  3. Shielding calculations for NET

    International Nuclear Information System (INIS)

    Verschuur, K.A.; Hogenbirk, A.

    1991-05-01

    In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab

  4. Benchmark analysis and evaluations of materials for shielding

    International Nuclear Information System (INIS)

    Benton, E.R.; Gersey, B.B.; Uchihori, Y.; Yasuda, N.; Kitamura, H.; Shavers, M.R.

    2005-01-01

    The goal of this project is to provide a benchmark set of heavy ion beam measurements behind ''standard'' targets made using radiation detectors routinely used for astronaut dosimetry and to test the radiation shielding properties of candidate multifunctional spacecraft materials. These measurements are used in testing and validating space radiation transport codes currently being developed by NASA and in selecting promising materials for further development. The radiation dosimetry instruments being used include CR-39 plastic nuclear track detector (PNTD), Tissue-Equivalent Proportional Counter (TEPC), the Liulin Mobile Dosimetry Unit (MDU) and thermoluminescent detector (TLD). Each set of measurements include LET/y spectra, and dose and dose equivalent as functions of shield thickness. Measurements are being conducted at the NIRS HIMAC, using heavy-ion beams of energy commonly encountered in the galactic cosmic ray (GCR) environment and that have been identified as being of particular concern to the radiation protection of space crews. Measurements are being made behind a set of standard'' targets including Al, Cu, polyethylene (HDPE) and graphite that vary in thickness from 0.5 to > 30 g/cm 2 . In addition, we are measuring the shielding properties of novel shielding materials being developed by and for NASA, including carbon and polymer composites. (author)

  5. Application of solid-phase micro extraction for the determination of pesticides in vegetable samples by gas chromatography with an electron capture detector

    International Nuclear Information System (INIS)

    Chai, Mee Kin; Tan, Guan Huat; Kumari, Asha

    2008-01-01

    A solid-phase micro extraction (SPME) method has been developed for the determination of 9 pesticides in 2 vegetables -cucumber and tomato - samples, based on direct immersion mode and subsequent desorption into the injection port of a gas chromatograph with an electron capture detector (GC-ECD). The main factors affecting the SPME process such as extraction time and temperature, desorption time and temperature, the effect of salt addition and fiber depth into the liner were studied and optimized. The analytical procedure proposed consisted of a 30 minute ultrasonic extraction of the target compounds from 1.0 g vegetable samples with 5 mL of distilled water. Then, the samples were filtered and topped up with distilled water to 10 mL. The analytes in this aqueous extract were extracted for 15 minutes with a 100 μm thickness polydimethylsiloxane SPME fiber. Relative standard deviations for triplicate analyses of samples were less than 10 %. The recoveries of the pesticides studied in cucumber and tomato ranged from 52 % to 82 % and the RSD were below 10 %. Therefore, the proposed method is applicable in the analysis of pesticides in vegetable matrices. SPME has been shown to be a simple extraction technique, which has a number of advantages such as solvent free extraction, simplicity and compatibility with the chromatographic analytical system. (author)

  6. Advances in the sample preparation and the detector for a combined solvent extraction-liquid scintillation method of low-level plutonium measurement

    International Nuclear Information System (INIS)

    Perdue, P.T.; Christian, D.J.; Thorngate, J.H.; McDowell, W.J.; Case, G.N.

    1976-07-01

    A combined solvent extraction-liquid scintillation technique, developed at Oak Ridge National Laboratory (ORNL), has many possible applications to the determination of low levels of plutonium and other alpha-emitting nuclides. Using these procedures, plutonium can be extracted from biological or environmental samples and introduced directly into a liquid scintillator. Quenching of the scintillator is thus minimized so that spectroscopic techniques may be employed. Existing chemical procedures and counting equipment were reviewed and improved. Purification of the di(2-ethylhexyl)phosphoric acid (used as the actinide extractant) was found necessary. Destruction of organic material in the sample and control of the valence state of plutonium were found to be major sources of irreproducibility. Methods were developed to allow samples separated with commonly used ion exchange techniques to be extracted into the scintillator. Comparisons were made of a wide variety of the components and parameters of the detector system to find the best combination of pulse-height resolution and pulse-shape discrimination. When a single phototube was used, optimum performance was obtained using a hemispherical reflector-sample holder viewed sideways by an RCA 8575 photomultiplier tube used in conjunction with a special integrating preamplifier and a good quality linear amplifier that used delay lines to shape the pulses

  7. Photon Shielding Features of Quarry Tuff

    Directory of Open Access Journals (Sweden)

    Vega-Carrillo Hector Rene

    2017-01-01

    Full Text Available Cantera is a quarry tuff widely used in the building industry; in this work the shielding features of cantera were determined. The shielding characteristics were calculated using XCOM and MCNP5 codes for 0.03, 0.07, 0.1, 0.3, 0.662, 1, 2, and 3 MeV photons. With XCOM the mass interaction coefficients, and the total mass attenuation coefficients, were calculated. With the MCNP5 code a transmission experiment was modelled using a point-like source located 42 cm apart from a point-like detector. Between the source and the detector, cantera pieces with different thickness, ranging from 0 to 40 cm were included. The collided and uncollided photon fluence, the Kerma in air and the Ambient dose equivalent were estimated. With the uncollided fluence the linear attenuation coefficients were determined and compared with those calculated with XCOM. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl-based γ-ray spectrometer and a 137Cs source.

  8. Monte Carlo simulations for the optimisation of low-background Ge detector designs

    Energy Technology Data Exchange (ETDEWEB)

    Hakenmueller, Janina; Heusser, Gerd; Maneschg, Werner; Schreiner, Jochen; Simgen, Hardy; Stolzenburg, Dominik; Strecker, Herbert; Weber, Marc; Westernmann, Jonas [Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, 69117 Heidelberg (Germany); Laubenstein, Matthias [Laboratori Nazionali del Gran Sasso, Via G. Acitelli 22, 67100 Assergi L' Aquila (Italy)

    2015-07-01

    Monte Carlo simulations for the low-background Ge spectrometer Giove at the underground laboratory of MPI-K, Heidelberg, are presented. In order to reduce the cosmogenic background at the present shallow depth (15 m w.e.) the shielding of the spectrometer includes an active muon veto and a passive shielding (lead and borated PE layers). The achieved background suppression is comparable to Ge spectrometers operated in much greater depth. The geometry of the detector and the shielding were implemented using the Geant4-based toolkit MaGe. The simulations were successfully optimised by determining the correct diode position and active volume. With the help of the validated Monte Carlo simulation the contribution of the single components to the overall background can be examined. This includes a comparison between simulated results and measurements with different fillings of the sample chamber. Having reproduced the measured detector background in the simulation provides the possibility to improve the background by reverse engineering of the passive and active shield layers in the simulation.

  9. Ionization detector

    International Nuclear Information System (INIS)

    Steele, D.S.

    1987-01-01

    An ionization detector having an array of detectors has, for example, grounding pads positioned in the spaces between some detectors (data detectors) and other detectors (reference detectors). The grounding pads are kept at zero electric potential, i.e. grounded. The grounding serves to drain away electrons and thereby prevent an unwanted accumulation of charge in the spaces, and cause the electric field lines to be more perpendicular to the detectors in regions near the grounding pads. Alternatively, no empty space is provided there being additional, grounded, detectors provided between the data and reference detectors. (author)

  10. Heavy metal oxide glasses as gamma rays shielding material

    International Nuclear Information System (INIS)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir

    2016-01-01

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal (_5_6Ba, _6_4Gd, _8_2Pb, _8_3Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  11. Heavy metal oxide glasses as gamma rays shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir, E-mail: dr.tejbir@gmail.com

    2016-10-15

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal ({sub 56}Ba, {sub 64}Gd, {sub 82}Pb, {sub 83}Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  12. A contribution to shielding effectiveness analysis of shielded tents

    Directory of Open Access Journals (Sweden)

    Vranić Zoran M.

    2004-01-01

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

  13. Two-dimensional shielding benchmarks for iron at YAYOI, (1)

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

    The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

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

    International Nuclear Information System (INIS)

    Souza, Gregorio Soares de

    2011-01-01

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

  15. Composites with carbon nanotube for radiation shielding application

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  16. Composites with carbon nanotube for radiation shielding application

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  17. A novel automatic flow method with direct-injection photometric detector for determination of dissolved reactive phosphorus in wastewater and freshwater samples.

    Science.gov (United States)

    Koronkiewicz, Stanislawa; Trifescu, Mihaela; Smoczynski, Lech; Ratnaweera, Harsha; Kalinowski, Slawomir

    2018-02-12

    The novel automatic flow system, direct-injection detector (DID) integrated with multi-pumping flow system (MPFS), dedicated for the photometric determination of orthophosphates in wastewater and freshwater samples is for the first time described. All reagents and the sample were injected simultaneously, in counter-current into the reaction-detection chamber by the system of specially selected for this purpose solenoid micro-pumps. The micro-pumps provided good precision and accuracy of the injected volumes. For the determination of orthophosphates, the molybdenum blue method was employed. The developed method can be used to detect orthophosphate in the range 0.1-12 mg L -1 , with the repeatability (RSD) about 2.2% at 4 mg L -1 and a very high injection throughput of 120 injections h -1 . It was possible to achieve a very small consumption of reagents (10 μL of ammonium molybdate and 10 μL of ascorbic acid) and sample (20 μL). The volume of generated waste was only 440 μL per analysis. The method has been successfully applied, giving a good accuracy, to determination of orthophosphates in complex matrix samples: treated wastewater, lake water and reference sample of groundwater. The developed system is compact, small in both size and weight, requires 12 V in supply voltage, which are desirable for truly portable equipment used in routine analysis. The simplicity of the system should result in its greater long-time reliability comparing to other flow methods previously described.

  18. Neutronic reactor thermal shield

    International Nuclear Information System (INIS)

    Lowe, P.E.

    1976-01-01

    A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

  19. A shield against distraction

    OpenAIRE

    Halin, N.; Marsh, J.E.; Hellman, A.; Hellstrom, I.; Sörqvist, Patrik

    2014-01-01

    In this paper, we apply the basic idea of a trade-off between the level of concentration and distractibility to test whether a manipulation of task difficulty can shield against distraction. Participants read, either in quiet or with a speech noise background, texts that were displayed either in an easy-to-read or a hard-to-read font. Background speech impaired prose recall, but only when the text was displayed in the easy-to-read font. Most importantly, recall was better in the background sp...

  20. Silicon detectors

    International Nuclear Information System (INIS)

    Klanner, R.

    1984-08-01

    The status and recent progress of silicon detectors for high energy physics is reviewed. Emphasis is put on detectors with high spatial resolution and the use of silicon detectors in calorimeters. (orig.)

  1. Measuring space radiation shielding effectiveness

    OpenAIRE

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

    2017-01-01

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

  2. Selective shielding device for scintiphotography

    International Nuclear Information System (INIS)

    Harper, J.W.; Kay, T.D.

    1976-01-01

    A selective shielding device to be used in combination with a scintillation camera is described. The shielding device is a substantially oval-shaped configuration removably secured to the scintillation camera. As a result of this combination scanning of preselected areas of a patient can be rapidly and accurately performed without the requirement of mounting any type of shielding paraphernalia on the patient. 1 claim, 2 drawing figures

  3. Tax Shield, Insolvenz und Zinsschranke

    OpenAIRE

    Arnold, Sven; Lahmann, Alexander; Schwetzler, Bernhard

    2010-01-01

    Dieser Beitrag analysiert den Wertbeitrag fremdfinanzierungsbedingter Steuervorteile (Tax Shield) unter realistischen Bedingungen (keine Negativsteuer; mögliche Insolvenz) für unterschiedliche Finanzierungspolitiken. Zusätzlich wird der Effekt der sogenannten Zinsschranke auf den Wert des Tax Shield ermittelt. Die Bewertung des Tax Shield mit und ohne Zinsschranke findet im einperiodigen Fall auf der Basis von Optionspreismodellen und im mehrperiodigen Fall auf der Basis von Monte Carlo Simul...

  4. SHIELD verification and validation report

    International Nuclear Information System (INIS)

    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

  5. Multifunctional Hot Structure Heat Shield

    Data.gov (United States)

    National Aeronautics and Space Administration — This project is performing preliminary development of a Multifunctional Hot Structure (HOST) heat shield for planetary entry. Results of this development will...

  6. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1978-01-01

    A shield for use with nuclear reactor systems to attenuate radiation resulting from reactor operation is described. The shield comprises a container preferably of a thin, flexible or elastic material, which may be in the form of a bag, a mattress, a toroidal segment or toroid or the like filled with radiation attenuating liuid. Means are provided in the container for filling and draining the container in place. Due to its flexibility, the shield readily conforms to irregularities in surfaces with which it may be in contact in a shielding position

  7. Testing of complementarity of PDA and MS detectors using chromatographic fingerprinting of genuine and counterfeit samples containing sildenafil citrate.

    Science.gov (United States)

    Custers, Deborah; Krakowska, Barbara; De Beer, Jacques O; Courselle, Patricia; Daszykowski, Michal; Apers, Sandra; Deconinck, Eric

    2016-02-01

    Counterfeit medicines are a global threat to public health. High amounts enter the European market, which is why characterization of these products is a very important issue. In this study, a high-performance liquid chromatography-photodiode array (HPLC-PDA) and high-performance liquid chromatography-mass spectrometry (HPLC-MS) method were developed for the analysis of genuine Viagra®, generic products of Viagra®, and counterfeit samples in order to obtain different types of fingerprints. These data were included in the chemometric data analysis, aiming to test whether PDA and MS are complementary detection techniques. The MS data comprise both MS1 and MS2 fingerprints; the PDA data consist of fingerprints measured at three different wavelengths, i.e., 254, 270, and 290 nm, and all possible combinations of these wavelengths. First, it was verified if both groups of fingerprints can discriminate between genuine, generic, and counterfeit medicines separately; next, it was studied if the obtained results could be ameliorated by combining both fingerprint types. This data analysis showed that MS1 does not provide suitable classification models since several genuines and generics are classified as counterfeits and vice versa. However, when analyzing the MS1_MS2 data in combination with partial least squares-discriminant analysis (PLS-DA), a perfect discrimination was obtained. When only using data measured at 254 nm, good classification models can be obtained by k nearest neighbors (kNN) and soft independent modelling of class analogy (SIMCA), which might be interesting for the characterization of counterfeit drugs in developing countries. However, in general, the combination of PDA and MS data (254 nm_MS1) is preferred due to less classification errors between the genuines/generics and counterfeits compared to PDA and MS data separately.

  8. The Tower Shielding Facility: Its glorious past

    International Nuclear Information System (INIS)

    Muckenthaler, F.J.

    1997-01-01

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

  9. 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.

  10. 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 ...

  11. SHIELDS Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Jordanova, Vania Koleva [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-03

    Predicting variations in the near-Earth space environment that can lead to spacecraft damage and failure, i.e. “space weather”, remains a big space physics challenge. A new capability was developed at Los Alamos National Laboratory (LANL) to understand, model, and predict Space Hazards Induced near Earth by Large Dynamic Storms, the SHIELDS framework. This framework simulates the dynamics of the Surface Charging Environment (SCE), the hot (keV) electrons representing the source and seed populations for the radiation belts, on both macro- and micro-scale. In addition to using physics-based models (like RAM-SCB, BATS-R-US, and iPIC3D), new data assimilation techniques employing data from LANL instruments on the Van Allen Probes and geosynchronous satellites were developed. An order of magnitude improvement in the accuracy in the simulation of the spacecraft surface charging environment was thus obtained. SHIELDS also includes a post-processing tool designed to calculate the surface charging for specific spacecraft geometry using the Curvilinear Particle-In-Cell (CPIC) code and to evaluate anomalies' relation to SCE dynamics. Such diagnostics is critically important when performing forensic analyses of space-system failures.

  12. Radiation shielding analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. Radiation Shielding Properties of Some Marbles in Turkey

    International Nuclear Information System (INIS)

    Guenoglu, K.; Akkurt, I.

    2011-01-01

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

  14. Radiation Shielding Properties of Some Marbles in Turkey

    Science.gov (United States)

    Günoǧlu, K.; Akkurt, I.

    2011-12-01

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

  15. Texture analysis with neutron bending on geological/mineralogical multi-phase samples using a locally resolving detector and profile analysis

    International Nuclear Information System (INIS)

    Merz, P.L.

    1991-02-01

    In the context of this work, the NANCY four circuit diffractometer of the University of Bonn at the RRJ2 research reactor at KFA Juelich was equipped with a linear locally resolving scintillation detector JULIOS. To evaluate the diffractogram occurring at a pole figure measurement, user-friendly profile analysis and other evaluation programs were developed on the PC. The course of evaluation was largely automated, so that only a few interactive steps are required. The measuring period of a sample is usually two to three days. Up to 35 pole figures are produced, depending on the phase conditions of the examined sample. The evaluation of up to 900 diffractograms with the aid of the automatically running profile analysis program takes between 30 and 100 minutes on a 20 MHz PC 386. Pole figure datafiles are produced from the intensity data obtained in this way by a conversion program. The texture analyses of copper pyrites ores introduced here are connected with geological questions. (orig.) [de

  16. Design and implementation of a simple on-line time-activity curve detector for [O-15] water PET studies

    International Nuclear Information System (INIS)

    Wollenweber, S.D.; Hichwa, R.D.; Ponto, L.L.B.

    1996-01-01

    A simple, automated on-line detector system has been fabricated and implemented to detect the arterial time-activity curve (TAC) for water PET studies. This system offers two significant improvements over existing systems: a pump mechanism is not required to control arterial blood flow through the detector and dispersion correction of the time-activity curve is unnecessary. The positrons emanating from a thin-walled, 0.134 cm inner-diameter plastic tube are detected by a 0.5 cm wide by 1.0 cm long by 0.1 cm thick plastic scintillator mounted to a miniature PMT. Photon background is shielded by a 2.0 cm thick cylindrical lead shield. Mean cerebral blood flow (mCBF) calculated from the TAC determined by 1-second automated sampling was compared to that calculated from every 5-second integrated manual samples. Improvements in timing resolution (1-sec vs. 5-sec) cause small but significant differences between the two sampling methods. Dispersion is minimized due to small tubing diameters, short lengths of tubing between the radial arterial sampling site and the detector and the presence of a 3-way valve 10 cm proximal to the detector

  17. Self shielding in cylindrical fissile sources in the APNea system

    International Nuclear Information System (INIS)

    Hensley, D.

    1997-01-01

    In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  19. Problems in radiation shielding calculations with Monte Carlo methods

    International Nuclear Information System (INIS)

    Ueki, Kohtaro

    1985-01-01

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

  20. Analysis of the JASPER Program Radial Shield Attenuation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1993-01-01

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

  1. Radiation shield for PWR reactors

    International Nuclear Information System (INIS)

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

    One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

  2. Nuclear data for radiation shielding

    International Nuclear Information System (INIS)

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

    The third shielding expert conference was convened in Paris in Oct. 1975 for exchanging informations about the sensitivity evaluation of nuclear data in shielding calculation and integral bench mark experiment. The requirements about nuclear data presented at present from the field of nuclear design do not reflect sufficiently the requirements of shielding design, therefore it was the object to gather the requirements about nuclear data from the field of shielding. The nuclides used for shielding are numerous, and the nuclear data on these isotopes are required. Some of them cannot be ignored as the source of secondary γ-ray or in view of the radioactivation of materials. The requirements for the nuclear data of neutrons in the field of shielding are those concerning the reaction cross sections producing secondary γ-ray, the reaction cross sections including the production of secondary neutrons, elastic scattering cross sections, and total cross sections. The topics in the Paris conference about neutron shielding data are described, such as the methodology of sensitivity evaluation, the standardization of group constant libraries, the bench mark experiment on iron and sodium, and the cross section of γ-ray production. In the shielding of nuclear fission reactors, the γ-ray production owing to nuclear fission reaction is also important. In (d, t) fusion reactors, high energy neutrons are generated, and high energy γ-ray is emitted through giant E1 resonance. (Kako, I.)

  3. Concrete shielding exterior to iron

    International Nuclear Information System (INIS)

    Yurista, P.; Cossairt, D.

    1983-08-01

    A rule of thumb at Fermilab has been to use 3 feet of concrete exterior to iron shielding. A recent design of a shield with a severe dimensional constraint has prompted a re-evaluation of this rule of thumb and has led to the following calculations of the concrete thickness required to nullify this problem. 4 references, 4 figures

  4. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    1975-06-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. This pamphlet will provide physicians and radiologic technologists with information which will aid their appropriate use of gonad shielding

  5. Study of Natural Radioactivity in Coal Samples of Baganuur Coal Mine, Mongolia

    Science.gov (United States)

    Altangerel, M.; Norov, N.; Altangerel, D.

    2009-03-01

    Coal and soil samples from Baganuur Coal Mine (BCM) of Mongolia have been investigated. The activities of 226Ra, 232Th and 40K have been measured by gamma-ray spectrometry using shielded HPGe detector. Contents of natural radionuclide elements (U, Th and K) have been determined. Also the activities and contents of radionuclide of ashes were determined which generated in Thermal Power Plant ♯3 of Ulaanbaatar from coal supplied from BCM.

  6. The Imperial Shield

    DEFF Research Database (Denmark)

    Mortensen, Simon Valentin

    2006-01-01

      The title of this Ph.d. dissertation is "The Imperial Shield: Imperial Overstretch, Assured Destruction, and the ban on nationwide ABM-defense with particular emphasis on the Johnson and the Nixon Administration". The dissertation set out to explain the origins of the ABM Treaty's central meaning....... Domestic spending continued to increase by more in real terms than the GDP, and the Democratically controlled Congress also made some very expensive modifications in Nixon tax bill in the fall of 1969, once again plunging the budget into the red.The economic crisis was partly caused by, and partly...... the Administration debated the deployment of new ABM-sites in early 1970, Kissinger could not prevail against these forces, but had to settle for a compromise, which he regarded as less than a definite commitment to nationwide ABM-defense.The political developments were of even greater importance. A strong link has...

  7. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  8. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  9. 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.

  10. Penetration shielding applications of CYLSEC

    International Nuclear Information System (INIS)

    Dexheimer, D.T.; Hathaway, J.M.

    1985-01-01

    Evaluation of penetration and discontinuity shielding is necessary to meet 10CFR20 regulations for ensuring personnel exposures are as low as reasonably achievable (ALARA). Historically, those shielding evaluations have been done to some degree on all projects. However, many early plants used conservative methods due to lack of an economical computer code, resulting in costly penetration shielding programs. With the increased industry interest in cost effectively reducing personnel exposures to meet ALARA regulations and with the development of the CYLSEC gamma transport computer code at Bechtel, a comprehensive effort was initiated to reduce penetration and discontinuity shielding but still provide a prudent degree of protection for plant personnel from radiation streaming. This effort was more comprehensive than previous programs due to advances in shielding analysis technology and increased interest in controlling project costs while maintaining personnel exposures ALARA. Methodology and resulting cost savings are discussed

  11. Shield calculations, optimization vs. paradigm

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  13. Parameters calculation of shielding experiment

    International Nuclear Information System (INIS)

    Gavazza, S.

    1986-02-01

    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) [pt

  14. Design experience: CRBRP radiation shielding

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  15. Semi-analytic flux formulas for shielding calculations

    International Nuclear Information System (INIS)

    Wallace, O.J.

    1976-06-01

    A special coordinate system based on the work of H. Ono and A. Tsuro has been used to derive exact semi-analytic formulas for the flux from cylindrical, spherical, toroidal, rectangular, annular and truncated cone volume sources; from cylindrical, spherical, truncated cone, disk and rectangular surface sources; and from curved and tilted line sources. In most of the cases where the source is curved, shields of the same curvature are allowed in addition to the standard slab shields; cylindrical shields are also allowed in the rectangular volume source flux formula. An especially complete treatment of a cylindrical volume source is given, in which dose points may be arbitrarily located both within and outside the source, and a finite cylindrical shield may be considered. Detector points may also be specified as lying within spherical and annular source volumes. The integral functions encountered in these formulas require at most two-dimensional numeric integration in order to evaluate the flux values. The classic flux formulas involving only slab shields and slab, disk, line, sphere and truncated cone sources become some of the many special cases which are given in addition to the more general formulas mentioned above

  16. Status of reactor-shielding research in the US

    International Nuclear Information System (INIS)

    Maienshein, F.C.

    1980-01-01

    While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study

  17. New Micromegas detectors in the CAST experiment

    International Nuclear Information System (INIS)

    Aune, S.; Braeuninger, H.; Dafni, T.; Fanourakis, G.; Ferrer Ribas, E.; Galan Lacarra, J.; Geralis, T.; Giomataris, I.; Iguaz, F.; Irastorza, I.G.; Kousouris, K.; Morales, J.; Mols, J.P.; Papaevangelou, T.; Pivovaroff, M.; Ruz, J.; Soufli, R.; Tomas, A.; Zachariadou, K.

    2009-01-01

    A low background Micromegas detector was operating at the sunrise side of the CERN Axion Solar Telescope (CAST) experiment during the previous data taking periods (2002-2006). This detector, constructed of low radioactivity materials, operated efficiently and achieved a background level, 5x10 -5 keV -1 cm -2 s -1 , in the 2-7 keV region. This performance was accomplished by exploiting the spatial and energy resolution of the detector as well as the time information contained in the pulse shape of the events. During the second phase of the experiment, the detector at the sunrise was replaced and upgraded by including a shielding. Moreover, the old time projection chamber (TPC) covering the sunset side of the experiment was replaced by two new Micromegas detectors. These detectors belong to the newest generation of Micromegas detectors: 'bulk' and 'microbulk'. Performances and advantages will be presented.

  18. Neutron self-shielding with k0-NAA irradiations

    International Nuclear Information System (INIS)

    Chilian, C.; Chambon, R.; Kennedy, G.

    2010-01-01

    A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

  19. Transparent fast neutron shielding material and shielding method

    International Nuclear Information System (INIS)

    Nashimoto, Tetsuji; Katase, Haruhisa.

    1993-01-01

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

  20. Study of gamma radiation shielding properties of ZnO-TeO_2 glasses

    International Nuclear Information System (INIS)

    Issa, Shama A.M.; Sayyed, M.I.; Kurudirek, Murat

    2017-01-01

    Mass attenuation coefficient (μm), half value layer (HVL) and mean free path (MFP) for xZnO-(100-x)TeO_2, where x=10, 15, 20, 25, 30, 35 and 40 mol%, have been measured for 0.662, 1.173 and 1.33 MeV photons emitted from "1"3"7Cs and "6"0Co using a 3 x 3 inch NaI (Tl) detector. Some relevant parameters such as effective atomic numbers (Z_e_f_f) and electron densities (Nel) of glass samples have been also calculated in the photon energy range of 0.015-15 MeV. Moreover, gamma-ray energy absorption buildup factor (EABF) and exposure buildup factor (EBF) were estimated using a five-parameter Geometric Progression (GP) fitting approximation, for penetration depths up to 40 MFP and in the energy range 0.015-15 MeV. The measured mass attenuation coefficients were found to agree satisfactorily with the theoretical values obtained through WinXcom. Effective atomic numbers (Z_e_f_f) and electron densities (N_e_l) were found to be the highest for 40ZnO-60TeO_2 glass in the energy range 0.04-0.2 MeV. The 10ZnO-90TeO_2 glass sample has lower values of gamma-ray EBFs in the intermediate energy region. The reported new data on radiation shielding characteristics of zinc tellurite glasses should be beneficial from the point of proper gamma shield designs when intended to be used as radiation shields. (author)

  1. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

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

  2. Transmutation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.c [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Lahodova, Z. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Klupak, V. [Nuclear Research Institute Rez plc (Czech Republic); Sus, F. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic); Kucera, J. [Research Centre Rez Ltd. (Czech Republic); Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic); Kus, P.; Marek, M. [Research Centre Rez Ltd. (Czech Republic); Nuclear Research Institute Rez plc (Czech Republic)

    2011-03-11

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

  3. Transmutation detectors

    International Nuclear Information System (INIS)

    Viererbl, L.; Lahodova, Z.; Klupak, V.; Sus, F.; Kucera, J.; Kus, P.; Marek, M.

    2011-01-01

    We have designed a new type of detectors, called transmutation detectors, which can be used primarily for neutron fluence measurement. The transmutation detector method differs from the commonly used activation detector method in evaluation of detector response after irradiation. Instead of radionuclide activity measurement using radiometric methods, the concentration of stable non-gaseous nuclides generated by transmutation in the detector is measured using analytical methods like mass spectrometry. Prospective elements and nuclear reactions for transmutation detectors are listed and initial experimental results are given. The transmutation detector method could be used primarily for long-term measurement of neutron fluence in fission nuclear reactors, but in principle it could be used for any type of radiation that can cause transmutation of nuclides in detectors. This method could also be used for measurement in accelerators or fusion reactors.

  4. Morphometry of terrestrial shield volcanoes

    Science.gov (United States)

    Grosse, Pablo; Kervyn, Matthieu

    2018-03-01

    Shield volcanoes are described as low-angle edifices built primarily by the accumulation of successive lava flows. This generic view of shield volcano morphology is based on a limited number of monogenetic shields from Iceland and Mexico, and a small set of large oceanic islands (Hawaii, Galápagos). Here, the morphometry of 158 monogenetic and polygenetic shield volcanoes is analyzed quantitatively from 90-meter resolution SRTM DEMs using the MORVOLC algorithm. An additional set of 24 lava-dominated 'shield-like' volcanoes, considered so far as stratovolcanoes, are documented for comparison. Results show that there is a large variation in shield size (volumes from 0.1 to > 1000 km3), profile shape (height/basal width (H/WB) ratios mostly from 0.01 to 0.1), flank slope gradients (average slopes mostly from 1° to 15°), elongation and summit truncation. Although there is no clear-cut morphometric difference between shield volcanoes and stratovolcanoes, an approximate threshold can be drawn at 12° average slope and 0.10 H/WB ratio. Principal component analysis of the obtained database enables to identify four key morphometric descriptors: size, steepness, plan shape and truncation. Hierarchical cluster analysis of these descriptors results in 12 end-member shield types, with intermediate cases defining a continuum of morphologies. The shield types can be linked in terms of growth stages and shape evolution, related to (1) magma composition and rheology, effusion rate and lava/pyroclast ratio, which will condition edifice steepness; (2) spatial distribution of vents, in turn related to the magmatic feeding system and the tectonic framework, which will control edifice plan shape; and (3) caldera formation, which will condition edifice truncation.

  5. Ice shielding in the large scale GENIUS experiment for double beta decay and dark matter search

    International Nuclear Information System (INIS)

    Klapdor-Kleingrothaus, H.V.; Zdesenko, Yu.G.

    1998-01-01

    We suggest here the use of ice as shielding material in the large scale GENIUS experiment for the ultimate sensitive double beta decay and dark matter search. The idea is to pack a working volume of several tons of liquid nitrogen, which contains the ''naked'' Ge detectors, inside an ice shielding. Very thin plastic foil would be used in order to prevent leakage of the liquid nitrogen. Due to the excellent advantages of ice shielding (high purity and low cost, self-supporting ability, thermo-isolation and optical properties, safety) this could be another possible way of realization of the GENIUS project. (orig.)

  6. A study of gamma shielding

    International Nuclear Information System (INIS)

    Roogtanakait, N.

    1981-01-01

    Gamma rays have high penetration power and its attenuation depends upon the thickness and the attenuation coefficient of the shield, so it is necessary to use the high density shield to attenuate the gamma rays. Heavy concrete is considered to be used for high radiation laboratory and the testing of the shielding ability and compressibility of various types of heavy concrete composed of baryte, hematite, ilmenite and galena is carried out. The results of this study show that baryte-ilmenite concrete is the most suitable for high radiation laboratory in Thailand

  7. Radiation protection/shield design

    International Nuclear Information System (INIS)

    Disney, R.K.

    1977-01-01

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

  8. The Micromegas detector of the CAST experiment

    International Nuclear Information System (INIS)

    Abbon, P; Andriamonje, S; Aune, S; Dafni, T; Davenport, M; Delagnes, E; Oliveira, R de; Fanourakis, G; Ribas, E Ferrer; Franz, J; Geralis, T; Giganon, A; Gros, M; Giomataris, Y; Irastorza, I G; Kousouris, K; Morales, J; Papaevangelou, T; Ruz, J; Zachariadou, K; Zioutas, K

    2007-01-01

    A low-background Micromegas detector has been operating in the CAST experiment at CERN for the search for solar axions during the first phase of the experiment (2002-2004). The detector, made out of low radioactivity materials, operated efficiently and achieved a very high level of background rejection (5 x 10 -5 counts keV -1 cm -2 s -1 ) without shielding

  9. Calibration of germanium detectors

    International Nuclear Information System (INIS)

    Bjurman, B.; Erlandsson, B.

    1985-01-01

    This paper describes problems concerning the calibration of germanium detectors for the measurement of gamma-radiation from environmental samples. It also contains a brief description of some ways of reducing the uncertainties concerning the activity determination. These uncertainties have many sources, such as counting statistics, full energy peak efficiency determination, density correction and radionuclide specific-coincidence effects, when environmental samples are investigated at close source-to-detector distances

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-11

    Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Thyroid shields and neck exposures in cephalometric radiography

    Directory of Open Access Journals (Sweden)

    Cunha-Cruz Joana

    2006-06-01

    Full Text Available Abstract Background The thyroid is among the more radiosensitive organs in the body. The goal of this study was twofold: (1 to evaluate age-related changes in what is exposed to ionizing radiation in the neck area, and (2 to assess thyroid shield presence in cephalometric radiographs Methods Cephalometric radiographs at one academic setting were sampled and neck exposure was related to calendar year and patient's gender and age. Results In the absence of shields, children have more vertebrae exposed than adults (p Conclusion In the absence of a thyroid shield, children have more neck structure exposed to radiation than adults. In agreement with other reports, thyroid shield utilization in this study was low, particularly in children.

  13. Radiation shielding member

    International Nuclear Information System (INIS)

    Nemezawa, Isao; Kimura, Tadahiro; Mizuochi, Akira; Omori, Tetsu

    1998-01-01

    A single body of a radiation shield comprises a bag prepared by welding or bonding a polyurethane sheet which is made flat while interposing metal plates at the upper and the lower portion of the bag. Eyelet fittings are disposed to the upper and the lower portions of the bag passing through the metal plates and the flat portion of the bag. Water supplying/draining ports are disposed to two upper and lower places of the bag at a height where the metal plates are disposed. Reinforcing walls welded or bonded to the inner wall surface of the bag are elongated in vertical direction to divide the inside of the bag to a plurality of cells. The bag is suspended and supported from a frame with S-shaped hooks inserted into the eyelet fittings as connecting means. A plurality of bags are suspended and supported from the frame at a required height by way of the eyelets at the lower portion of the suspended and supported bag and the eyelet fittings at the upper portion of the bag below the intermediate connection means. (I.N.)

  14. Self-shielding factors

    International Nuclear Information System (INIS)

    Kaul, D.C.

    1982-01-01

    Throughout the last two decades many efforts have been made to estimate the effect of body self-shielding on organ doses from externally incident neutrons and gamma rays. These began with the use of simple geometry phantoms and have culminated in the use of detailed anthropomorphic phantoms. In a recent effort, adjoint Monte Carlo analysis techniques have been used to determine dose and dose equivalent to the active marrow as a function of energy and angle of neutron fluence externally incident on an anthropomorphic phantom. When combined with fluences from actual nuclear devices, these dose-to-fluence factors result in marrow dose values that demonstrate great sensitivity to variations in device type, range, and body orientation. Under a state-of-the-art radiation transport analysis demonstration program for the Japanese cities, sponsored by the Defense Nuclear Agency at the request of the National Council on Radiation Protection and Measurements, the marrow dose study referred to above is being repeated to obtain spectral distributions within the marrow for externally incident neutrons and gamma rays of arbitrary energy and angle. This is intended to allow radiobiologists and epidemiologists to select and to modify numbers of merit for correlation with health effects and to permit a greater understanding of the relationship between human and laboratory subject dosimetry

  15. Shielding plug device

    International Nuclear Information System (INIS)

    Orii, Shoichi; Hasegawa, Satoshi; Makishima, Kenji.

    1976-01-01

    Object: To reduce the size of and extend the life of a revolving bearing and facilitate the laying of driving cables and duct lines, this being accomplished by providing plug raising means of a fast breeder on a stationary plug mounting base so as to prevent the shearing force of sodium from acting upon the revolving bearing. Structure: The shield plug means comprises a stationary plug secured to the open end of the reactor container, a rotary plug rotatable with respect to the stationary plug, an annular base formed on top of the stationary plug so as to cover the rotary plug, a bearing secured to the rotary plug edge lower face and upper and lower locking plates. At the time of the rotation of the rotary plug, the upper locking plate is withdrawn, the stationary plug is raised to release the seal structure, and the lower locking plate is inserted between the bearing and stationary plug. In this way, smooth rotation of the rotary plug can be obtained. (Horiuchi, T.)

  16. Automatic modeling using PENELOPE of two HPGe detectors used for measurement of environmental samples by γ-spectrometry from a few sets of experimental efficiencies

    Science.gov (United States)

    Guerra, J. G.; Rubiano, J. G.; Winter, G.; Guerra, A. G.; Alonso, H.; Arnedo, M. A.; Tejera, A.; Mosqueda, F.; Martel, P.; Bolivar, J. P.

    2018-02-01

    The aim of this paper is to characterize two HPGe gamma-ray detectors used in two different laboratories for environmental radioactivity measurements, so as to perform efficiency calibrations by means of Monte Carlo Simulation. To achieve such an aim, methodologies developed in previous papers have been applied, based on the automatic optimization of the model of detector, so that the differences between computational and reference FEPEs are minimized. In this work, such reference FEPEs have been obtained experimentally from several measurements of the IAEA RGU-1 reference material for specific source-detector arrangements. The models of both detectors built through these methodologies have been validated by comparing with experimental results for several reference materials and different measurement geometries, showing deviations below 10% in most cases.

  17. Ionizing radiation detector using multimode optical fibers

    International Nuclear Information System (INIS)

    Suter, J.J.; Poret, J.C.; Rosen, M.; Rifkind, J.M.

    1993-01-01

    An optical ionizing radiation detector, based on the attenuation of 850-nm light in 50/125-μm multimode fibers, is described. The detector is especially well suited for application on spacecraft because of its small design. The detection element consists of a section of coiled fibers that has been designed to strip higher-order optical modes. Cylindrical radiation shields with atomic numbers ranging from Z = 13 (aluminum too) Z = 82 (lead) were placed around the ionizing radiation detector so that the effectiveness of the detector could be measured. By exposing the shields and the detector to 1.25-MeV cobalt 60 radiation, the mass attenuation coefficients of the shields were measured. The detector is based on the phenomenon that radiation creates optical color centers in glass fibers. Electron spin resonance spectroscopy performed on the 50/125-μm fibers showed the presence of germanium oxide and phosphorus-based color centers. The intensity of these centers is directly related to the accumulated gamma radiation

  18. On active shieldings in (ββ)0ν 76Ge decay experiments

    International Nuclear Information System (INIS)

    Garcia, E.; Morales, A.; Morales, J.; Nunez-Lagos, R.; Ortiz de Solorzano, A.; Puimedon, J.; Saenz, C.; Salinas, A.; Sarsa, M.L.; Villar, J.A.

    1992-01-01

    The sensitivity of an ultra low background Ge detector for the (ββ) 0ν decay of 76 Ge is estimated in two different experimental set-ups. The main difference between them is the inclusion or not of an active NaI shielding. We find that sensitivity of the Ge detector is not improved by this active shielding either for the O + -->O + or the O + -->2 + (ββ) 0ν transitions. Our results provide a valuable information for future 76 Ge enriched experiments. (orig.)

  19. Response of combined albedo-track neutron personnel dosimeters behind IHEP proton synchrotron shielding

    International Nuclear Information System (INIS)

    Sannikov, A.V.; Korshunova, E.P.

    1989-01-01

    The method of readings interpretation of combined albedo-track neutron personnel dosemeters based on calculationsl analysis of the detector responses in various neutron spectra is described. The measurements of dose equivalent responses have been performed in various points behind IHEP proton synchrotron shielding. It is shown that CDs with fission track detectors have a small dose equivalent response dispersion behind IHEP proton synchrotron shielding, that shows the promise of their using for neutron personnel monitoring, that shows the promise of their using for neutron personnel monitoring at high energy accelerators. 16 refs.; 7 figs.; 3 tabs

  20. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  1. Active Radiation Shield, Phase I

    Data.gov (United States)

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

  2. Gonad shielding in computerized tomography

    International Nuclear Information System (INIS)

    Rockstroh, G.

    1984-01-01

    The reduction of gonadal dose by shielding of the gonads was investigated for a Somatom 2 using an anthropomorphic phantom. For small distances from the slice examined the gonadal dose results from intracorporal secondary radiation and is only insignificantly reduced by shielding. For greater distances shielding is relatively more effective, the gonadal dose however is small because of the approximately exponential decay. Shielding of the gonads therefore does not seem adequate for the reduction of gonadal dose. From dose measurements in cylinder phantoms of several diameters it appears that no different results would be obtained for children and young adults. An effective reduction of gonadal dose is only possible with lead capsules for males. (author)

  3. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2005-01-01

    Full text: 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 120mm 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. (authors)

  4. Shielding calculations. Optimization vs. Paradigms

    International Nuclear Information System (INIS)

    Cornejo Diaz, Nestor; Hernandez Saiz, Alejandro; Martinez Gonzalez, Alina

    2005-01-01

    Many radiation shielding barriers in Cuba have been designed according to the criterion of Maxi-mum Projected Dose Rates. This fact has created the paradigm of low dose rates. Because of this, dose rate levels greater than units of Sv.h-1 would be considered unacceptable by many specialists, regardless of the real exposure times. Nowadays many shielding barriers are being designed using dose constraints in real exposure times. Behind the new barriers, dose rates could be notably greater than those behind the traditional ones, and it does not imply inadequate designs or constructive errors. In this work were obtained significant differences in dose rate levels and shield-ing thicknesses calculated by both methods for some typical installations. The work concludes that real exposure time approach is more adequate in order to optimise Radiation Protection, although this method should be carefully applied

  5. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2006-01-01

    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

  6. Detector Unit

    CERN Multimedia

    1960-01-01

    Original detector unit of the Instituut voor Kernfysisch Onderzoek (IKO) BOL project. This detector unit shows that silicon detectors for nuclear physics particle detection were already developed and in use in the 1960's in Amsterdam. Also the idea of putting 'strips' onto the silicon for high spatial resolution of a particle's impact on the detector were implemented in the BOL project which used 64 of these detector units. The IKO BOL project with its silicon particle detectors was designed, built and operated from 1965 to roughly 1977. Detector Unit of the BOL project: These detectors, notably the ‘checkerboard detector’, were developed during the years 1964-1968 in Amsterdam, The Netherlands, by the Natuurkundig Laboratorium of the N.V. Philips Gloeilampen Fabrieken. This was done in close collaboration with the Instituut voor Kernfysisch Onderzoek (IKO) where the read-out electronics for their use in the BOL Project was developed and produced.

  7. Detector trends

    International Nuclear Information System (INIS)

    Charpak, G.

    1986-01-01

    The author describes briefly the development of detectors for high energy physics experiments. Especially considered are semiconductor microstrip detectors, drift tubes, holographic bubble chambers, scintillating fiber optics, and calorimeters. (HSI).

  8. Infrared detectors

    CERN Document Server

    Rogalski, Antonio

    2010-01-01

    This second edition is fully revised and reorganized, with new chapters concerning third generation and quantum dot detectors, THz detectors, cantilever and antenna coupled detectors, and information on radiometry and IR optics materials. Part IV concerning focal plane arrays is significantly expanded. This book, resembling an encyclopedia of IR detectors, is well illustrated and contains many original references … a really comprehensive book.-F. Sizov, Institute of Semiconductor Physics, National Academy of Sciences, Kiev, Ukraine

  9. Rapid detailed characterization of concrete shielding blocks utilizing internal natural radionuclides for calibration

    International Nuclear Information System (INIS)

    McDonald, R.J.; Smith, A.R.; Hurley, D.L.; Norman, E.B.; Schoonover, M.R.

    1998-01-01

    Following many years of productive research, the 184-inch Cyclotron, the SuperHILAC, and the BEVALAC accelerators at the Berkeley Laboratory were closed, leaving thousands of concrete shielding blocks available for reuse, recycling, or disposal. The process history of these blocks precludes free release pending radiological characterization. This paper describes a procedure whereby a high efficiency shielded germanium spectrometer is used to rapidly characterize natural and man-made activity within the blocks. The spectrometer is moved up to the block and 5 minutes of data are collected at the point on the block that registers highest on a micro-R meter. Sensitivity is better than 1 pCi/g (0.037 Bq/g) for Co-60 and Eu-152, the prominent man-made activities observed. One-time calibration of the detector system is obtained from a sample of concrete, drilled with a hammer drill, counted in our low-background facility, and compared to crushed rock with known U, Th, and K activity. A simple relationship exists between the counts/minute observed in a characteristic gamma-ray peak and the activity in the block. (author)

  10. Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

    Energy Technology Data Exchange (ETDEWEB)

    McArthur, Matthew S., E-mail: matthew.s.mcarthur@gmail.com; Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu; Czirr, J. Bart, E-mail: czirr@juno.com

    2016-08-11

    Using the combination of a neutron-sensitive {sup 6}Li glass scintillator detector with a neutron-insensitive {sup 7}Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on {sup 6}Li. We used this detector with a {sup 252}Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

  11. Studying the shielding properties of lead glass composites using neutrons and gamma rays

    International Nuclear Information System (INIS)

    Osman, A.M.; El-Sarraf, M.A.; Abdel-Monem, A.M.; El-Sayed Abdo, A.

    2015-01-01

    Highlights: • Samples of sodalime silica glass loaded with different ratios of PbO were prepared. • Leaded glass composites were investigated for radiation shielding. • Experimental and theoretical attenuation parameters were studied. • Experimental and theoretical (MCNP5) results were in good agreement. - Abstract: The present work deals with the shielding properties of lead glass composites to find out its integrity for practical shielding applications and radiological safety. Composites of different lead oxide ratios (x = 0, 5, 10, 15 and 25 wt.%) have been prepared by the Nasser Glass and Crystal Company (Egypt). Attenuation measurements have been carried out using a collimated emitted beam from a fission 252 Cf (100 μg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D.) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes were measured using the BF3 detector and thermal neutron detection system. The attenuation relations were used to evaluate fast neutron macroscopic effective removal cross-section Σ R-Meas (cm −1 ), gamma rays total attenuation coefficient μ (cm −1 ) and thermal neutron macroscopic cross-section Σ Meas (cm −1 ). Theoretical calculations have been achieved using MCNP5 code to calculate the same two parameters. Also, MERCSF-N program was used to calculate fast neutron macroscopic removal cross-section Σ R-MER (cm −1 ). Measured and MCNP5 calculated results have been compared and were found to be in reasonable agreement

  12. Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

    Science.gov (United States)

    DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

    2018-06-01

    A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

  13. GARLIC, a shielding program for GAmma Radiation from Line- and Cylinder-sources

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-07-15

    GARLIC is a program for computing the gamma ray flux or dose rate at a shielded idotropic point detector, due to a line source or the line equivalent of a cylindrical source. The source strength distribution along the line must be either uniform or an arbitrary part of the positive half-cycle of a cosine function. The line source can be oriented arbitrarily with respect to the main shield and the detector, except that the detector must not be located on the line source or on its extension. The main source is a homogeneous plane slab in which scattered radiation is accounted for by multiplying each point element of the line source by a point source build-up factor inside the integral over the point elements. Between, the main shield and the line source additional shields can be introduced, which are either plane slabs, parallel to the main shield, or cylindrical rings, coaxial with the line source. Scattered radiation in the additional shields can only be accounted for by constant build-up factors outside the integral. GARLIC-xyz is an extended version particularly suited for the frequently met problem of shielding a room containing a large number of line sources in different positions. The program computes the angles and linear dimensions of a problem for GARLIC when the positions of the detector point and the end points of the line source are given as points in an arbitrary rectangular coordinate system. As an example the isodose curves in water are presented for a monoenergetic cosine-distributed line source at several source energies and for an operating fuel element of the Swedish reactor R3.

  14. Full energy peak efficiency of composite detectors for high energy gamma-rays

    International Nuclear Information System (INIS)

    Kshetri, Ritesh

    2015-01-01

    Experiments involving radioactive beams demand high detection efficiencies. One of the ways to obtain high detection efficiency without deteriorating the energy resolution or timing characteristics is the use of composite detectors which are composed of standard HPGe crystals arranged in a compact way. Two simplest composite detectors are the clover and cluster detectors. The TRIUMF-ISAC Gamma-Ray Escape-Suppressed Spectrometer (TIGRESS) comprises of 16 large volume, 32-fold segmented HPGe clover detectors, where each detector is shielded by a 20-fold segmented escape suppression shield (ESS)

  15. A variable temperature cryostat that produces in situ clean-up germanium detector surfaces

    International Nuclear Information System (INIS)

    Pehl, R.H.; Madden, N.W.; Malone, D.F.; Cork, C.P.; Landis, D.A.; Xing, J.S.; Friesel, D.L.

    1988-11-01

    Variable temperature cryostats that can maintain germanium detectors at temperatures from 82 K to about 400 K while the thermal shield surrounding the detectors remains much colder when the detectors are warmed have been developed. Cryostats such as these offer the possibility of cryopumping material from the surface of detectors to the colder thermal shield. The diode characteristics of several detectors have shown very significant improvement following thermal cycles up to about 150 K in these cryostats. Important applications for cryostats having this attribute are many. 4 figs

  16. Gravitational Field Shielding by Scalar Field and Type II Superconductors

    Directory of Open Access Journals (Sweden)

    Zhang B. J.

    2013-01-01

    Full Text Available The gravitational field shielding by scalar field and type II superconductors are theoret- ically investigated. In accord with the well-developed five-dimensional fully covariant Kaluza-Klein theory with a scalar field, which unifies the Einsteinian general relativity and Maxwellian electromagnetic theory, the scalar field cannot only polarize the space as shown previously, but also flatten the space as indicated recently. The polariza- tion of space decreases the electromagnetic field by increasing the equivalent vacuum permittivity constant, while the flattening of space decreases the gravitational field by decreasing the equivalent gravitational constant. In other words, the scalar field can be also employed to shield the gravitational field. A strong scalar field significantly shield the gravitational field by largely decreasing the equivalent gravitational constant. According to the theory of gravitational field shielding by scalar field, the weight loss experimentally detected for a sample near a rotating ceramic disk at very low tempera- ture can be explained as the shielding of the Earth gravitational field by the Ginzburg- Landau scalar field, which is produced by the type II superconductors. The significant shielding of gravitational field by scalar field produced by superconductors may lead to a new spaceflight technology in future.

  17. Thyroid shields and neck exposures in cephalometric radiography

    International Nuclear Information System (INIS)

    Hujoel, Philippe; Hollender, Lars; Bollen, Anne-Marie; Young, John D; Cunha-Cruz, Joana; McGee, Molly; Grosso, Alex

    2006-01-01

    The thyroid is among the more radiosensitive organs in the body. The goal of this study was twofold: (1) to evaluate age-related changes in what is exposed to ionizing radiation in the neck area, and (2) to assess thyroid shield presence in cephalometric radiographs Cephalometric radiographs at one academic setting were sampled and neck exposure was related to calendar year and patient's gender and age. In the absence of shields, children have more vertebrae exposed than adults (p < 0.0001) and females have more neck tissue exposed inferior to the hyoid bone than males (p < 0.0001). The hyoid bone-porion distance increased with age (p <0.01). Thyroid shields were visible in 19% of the radiographs and depended strongly on the calendar year during which patient was seen (p-value <0.0001). Compared to adults, children were less likely to wear thyroid shields, particularly between 1973 and 1990 (1.8% versus 7.3% – p-value < 0.05) and between 2001 and 2003 (7.1% versus 42.9% – p-value < 0.05). In the absence of a thyroid shield, children have more neck structure exposed to radiation than adults. In agreement with other reports, thyroid shield utilization in this study was low, particularly in children

  18. Development of a technique for the efficiency calibration of a HPGe detector for the off gas samples of a nuclear reactor

    International Nuclear Information System (INIS)

    Singh, Sarbjit; Agarwal, Chhavi; Ramaswami, A.; Manchanda, V.K.

    2007-01-01

    Regular monitoring of off gases released to the environment from a nuclear reactor is mandatory. The gaseous fission products are estimated by gamma ray spectrometry using a HPGe detector coupled to a multichannel analyser. In view of the lack of availability of gaseous fission products standards, an indirect method based on the charcoal absorption technique was developed for the efficiency calibration of HPGe detector system using 133B a and 152E u standards. The known activities of 133B a and 152E u are uniformly distributed in a vial having activated charcoal and counted on the HPGe detector system at liquid nitrogen temperature to determine the gamma ray efficiency for the vial having activated charcoal. The ratio of the gamma ray efficiencies of off gas present in the normal vial and the vial having activated charcoal at liquid nitrogen temperature are used to determine the gamma ray efficiency of off gas present in the normal vial. (author)

  19. Electrodynamic Dust Shield Demonstrator

    Science.gov (United States)

    Stankie, Charles G.

    2013-01-01

    The objective of the project was to design and manufacture a device to demonstrate a new technology developed by NASA's Electrostatics and Surface Physics Laboratory. The technology itself is a system which uses magnetic principles to remove regolith dust from its surface. This project was to create an enclosure that will be used to demonstrate the effectiveness of the invention to The Office of the Chief Technologist. ONE of the most important challenges of space exploration is actually caused by something very small and seemingly insignificant. Dust in space, most notably on the moon and Mars, has caused many unforeseen issues. Dirt and dust on Earth, while a nuisance, can be easily cleaned and kept at bay. However, there is considerably less weathering and erosion in space. As a result, the microscopic particles are extremely rough and abrasive. They are also electrostatically charged, so they cling to everything they make contact with. This was first noted to be a major problem during the Apollo missions. Dust would stick to the spacesuits, and could not be wiped off as predicted. Dust was brought back into the spacecraft, and was even inhaled by astronauts. This is a major health hazard. Atmospheric storms and other events can also cause dust to coat surfaces of spacecraft. This can cause abrasive damage to the craft. The coating can also reduce the effectiveness of thermal insulation and solar panels.' A group of engineers at Kennedy Space Center's Electrostatics and Surface Physics Laboratory have developed a new technology, called the Electrodynamic Dust Shield, to help alleviate these problems. It is based off of the electric curtain concept developed at NASA in 1967. "The EDS is an active dust mitigation technology that uses traveling electric fields to transport electrostatically charged dust particles along surfaces. To generate the traveling electric fields, the EDS consists of a multilayer dielectric coating with an embedded thin electrode grid

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  1. A large solid angle multiparameter neutron detector

    International Nuclear Information System (INIS)

    Ricco, G.; Anghinolfi, M.; Corvisiero, P.; Durante, E.; Maggiolo, S.; Prati, P.; Rottura, A.; Taiuti, M.

    1991-01-01

    A 4π neutron detector has been realized using organic scintillators: the detector is suitable for high efficiency, low background measurements of very low neutron rates in the 0.6-5 MeV energy range. Gamma-neutron discrimination has been performed by pulse shape, energy and neutron lifetime analysis and backgrounds have been reduced by anticoincidence detectors and paraffin-lead shielding. Tests of efficiency, energy resolution and radiation identification have been made with a low intensity Am-Be neutron source. (orig.)

  2. Radiation shielding in dental radiography

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  4. Shielding features of quarry stone

    International Nuclear Information System (INIS)

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

    2010-10-01

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

  5. 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.

  6. MMW [multimegawatt] shielding design and analysis

    International Nuclear Information System (INIS)

    Olson, A.P.

    1988-01-01

    Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

  7. Magnetic shielding for superconducting RF cavities

    Science.gov (United States)

    Masuzawa, M.; Terashima, A.; Tsuchiya, K.; Ueki, R.

    2017-03-01

    Magnetic shielding is a key technology for superconducting radio frequency (RF) cavities. There are basically two approaches for shielding: (1) surround the cavity of interest with high permeability material and divert magnetic flux around it (passive shielding); and (2) create a magnetic field using coils that cancels the ambient magnetic field in the area of interest (active shielding). The choice of approach depends on the magnitude of the ambient magnetic field, residual magnetic field tolerance, shape of the magnetic shield, usage, cost, etc. However, passive shielding is more commonly used for superconducting RF cavities. The issue with passive shielding is that as the volume to be shielded increases, the size of the shielding material increases, thereby leading to cost increase. A recent trend is to place a magnetic shield in a cryogenic environment inside a cryostat, very close to the cavities, reducing the size and volume of the magnetic shield. In this case, the shielding effectiveness at cryogenic temperatures becomes important. We measured the permeabilities of various shielding materials at both room temperature and cryogenic temperature (4 K) and studied shielding degradation at that cryogenic temperature.

  8. Superconducting magnetic shields production. Realisation d'ecrans magnetiques supraconducteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lainee, F; Kormann, R [Thomson-CSF, Domaine de Corbeville, 91 - Orsay (FR); Lainee, F [Ecole des Mines de Paris, 91 - Evry (FR)

    1992-02-01

    Low fields and low frequency shielding properties of YBCO magnetic shields are measured at 77 K. They compare favourably with shielding properties of mumetal shields. Therefore high-T{sub c} superconducting magnetic shields can already be used to shield small volumes. The case of magnetic shields for large volumes is also discussed. 3 refs; 6 figs; 4 tabs.

  9. Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule.

    Science.gov (United States)

    Pugliese, M; Bengin, V; Casolino, M; Roca, V; Zanini, A; Durante, M

    2010-08-01

    Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.

  10. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  11. Shield cost minimization using SWAN

    International Nuclear Information System (INIS)

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

    1993-01-01

    The common approach to the search for minimum cost shield designs is open-quotes trial-and-errorclose quotes; it proceeds as follows: 1. Based on prior experience and intuition, divide the shield into zones and assume their composition. 2. Solve the transport equation and calculate the relevant performance characteristics. 3. Change the composition or the geometry of one or a few of the zones and repeat step 2. 4. Repeat step 3 many times until the shield design appears to be optimal. 5. Select a different set of constituents and repeat steps 2,3, and 4. 6. Repeate step 5 a few or many times until the designer can point to the most cost-effective design

  12. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1980-01-01

    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)

  13. Determination of gamma ray shielding parameters of rocks and concrete

    Science.gov (United States)

    Obaid, Shamsan S.; Gaikwad, Dhammajyot K.; Pawar, Pravina P.

    2018-03-01

    Gamma shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff) and electron density (Neff) have been measured and calculated for rocks and concrete in the energy range 122-1330 keV. The measurements have been carried out at 122, 356, 511, 662, 1170, 1275, 1330 keV gamma ray energies using a gamma spectrometer includes a NaI(Tl) scintillation detector and MCA card. The atomic and electronic cross sections have also been investigated. Experimental and calculated (WinXCom) values were compared, and good agreement has been observed within the experimental error. The obtained results showed that feldspathic basalt, compact basalt, volcanic rock, dolerite and pink granite are more efficient than the sandstone and concrete for gamma ray shielding applications.

  14. Calculation of coincidence summing corrections for a specific small soil sample geometry

    Energy Technology Data Exchange (ETDEWEB)

    Helmer, R.G.; Gehrke, R.J.

    1996-10-01

    Previously, a system was developed at the INEL for measuring the {gamma}-ray emitting nuclides in small soil samples for the purpose of environmental monitoring. These samples were counted close to a {approx}20% Ge detector and, therefore, it was necessary to take into account the coincidence summing that occurs for some nuclides. In order to improve the technical basis for the coincidence summing corrections, the authors have carried out a study of the variation in the coincidence summing probability with position within the sample volume. A Monte Carlo electron and photon transport code (CYLTRAN) was used to compute peak and total efficiencies for various photon energies from 30 to 2,000 keV at 30 points throughout the sample volume. The geometry for these calculations included the various components of the detector and source along with the shielding. The associated coincidence summing corrections were computed at these 30 positions in the sample volume and then averaged for the whole source. The influence of the soil and the detector shielding on the efficiencies was investigated.

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

    Science.gov (United States)

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

    2018-05-01

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

  16. Calculation of the BREN house shielding experiments

    International Nuclear Information System (INIS)

    Woolson, William A.; Gritzner, Michael L.

    1987-01-01

    The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

  17. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    1993-05-01

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

  18. Nuclear steam generator tubesheet shield

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  19. Incorporation of Photon Analysis into an Active Interrogation System for Shielded Uranium Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Canion, Bonnie E. [Univ. of Texas, Austin, TX (United States)

    2016-02-01

    The main goal of this project is to investigate how photon and neutron signatures from an Associated Particle Imaging (API) Deuterium-Tritium (DT) neutron generator detector system can be used to non-destructively predict the enrichment of uranium in an unknown configuration of shielded uranium.

  20. Optimized Shielding and Fabrication Techniques for TiN and Al Microwave Resonators

    Science.gov (United States)

    Kreikebaum, John Mark; Kim, Eunseong; Livingston, William; Dove, Allison; Calusine, Gregory; Hover, David; Rosenberg, Danna; Oliver, William; Siddiqi, Irfan

    We present a systematic study of the effects of shielding and packaging on the internal quality factor (Qi) of Al and TiN microwave resonators designed for use in qubit readout. Surprisingly, Qi =1.3x106 TiN samples investigated at 100 mK exhibited no significant changes in linewidth when operated without magnetic shielding and in an open cryo-package. In contrast, Al resonators showed systematic improvement in Qi with each successive shield. Measurements were performed in an adiabatic demagnetization refrigerator, where typical ambient fields of 0.2 mT are present at the sample stage. We discuss the effect of 100 mK and 500 mK Cu radiation shields and cryoperm magnetic shielding on resonator Q as a function of temperature and input power in samples prepared with a variety of surface treatments, fabrication recipes, and embedding circuits. This research was supported by the ARO and IARPA.

  1. Detection limits of the NaI(Tl) shielded HPGe spectrometer

    International Nuclear Information System (INIS)

    Bikit, I.; Slivka, J.; Krmar, M.; Durcic, Z.; Zikic, N.; Conkic, Lj.; Veskovic, M.; Anicin, I.

    1999-01-01

    The results of a detailed study of the low-level performance of a NaI(Tl) shield added to an iron shielded HPGe spectrometer are presented. Both the 'slow' and the 'fast' anticoincidence gating modes were tested, the 'slow' mode being found better suited for general low-level spectroscopy applications. In long runs the stability of the system in this mode is satisfactory. The anticoincidence action of the NaI(T1) shield lowers the integral background of the iron shielded HPGe detector in the energy range from 30 keV to 2 MeV by a factor of 6.5, and suppresses the continuum above 150 keV by a factor larger than 10

  2. Shielding structure analysis for LSDS facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  3. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Saito, Tetsuo

    1983-01-01

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

  4. Shielding structure analysis for LSDS facility

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Terrestrial Background Reduction in RPM Systems by Direct Internal Shielding

    International Nuclear Information System (INIS)

    Robinson, Sean M.; Ashbaker, Eric D.; Schweppe, John E.

    2008-01-01

    Gamma-ray detection systems that are close to the earth or other sources of background radiation often require shielding, especially when trying to detect a relatively weak source. One particular case of interest that we address in this paper is that encountered by the Radiation Portal Monitors (RPMs) systems placed at border-crossing Ports of Entry (POE). These RPM systems are used to screen for illicit radiological materials, and they are often placed in situations where terrestrial background is large. In such environments, it is desirable to consider simple physical modifications that could be implemented to reduce the effects from background radiation without affecting the flow of traffic and the normal operation of the portal. Simple modifications include adding additional shielding to the environment, either inside or outside the apparatus. Previous work (2) has shown the utility of some of these shielding configurations for increasing the Signal to Noise Ratio (SNR) of gross-counting RPMs. Because the total cost for purchasing and installing RPM systems can be quite expensive, in the range of hundreds of thousands of dollars for each cargo-screening installation, these shielding variations may offer increases in detection capability for relatively small cost. Several modifications are considered here in regard to their real-world applicability, and are meant to give a general idea of the effectiveness of the schemes used to reduce background for both gross-counting and spectroscopic detectors. These scenarios are modeled via the Monte-Carlo N-Particle (MCNP) code package (1) for ease of altering shielding configurations, as well as enacting unusual scenarios prior to prototyping in the field. The objective of this paper is to provide results representative of real modifications that could enhance the sensitivity of this, as well as the next generation of radiation detectors. The models used in this work were designed to provide the most general results for

  6. An automated measuring system based on gamma spectrometry with HPGe detectors

    International Nuclear Information System (INIS)

    Mala, Helena; Rulik, Petr; Hyza, Miroslav; Dragounova, Lenka; Helebrant, Jan; Hroznicek, Marek; Jelinek, Pavel; Zak, Jan

    2016-01-01

    An automatic system for unattended gamma spectrometric measurements of bulk samples ( “Gamma Automat”, GA) was developed by the National Radiation Protection Institute and Nuvia, Inc. as a part of a research project. The basic parts include a detection system with two HPGe detectors in the lead shielded chambers, sample changer, sample tray and a control unit. The GA enables counting in two geometries: (i) with cylindrical containers (200 ml) either one placed at the detector face or 2-6 placed around the detector or (II) with Marinelli beakers (600 ml). The shelf can accommodate 180 cylindrical containers or 54 Marinelli beakers. Samples are changed by a robotic arm. The sample data and the analysis required are passed to the GA by a matrix code (generated within the laboratory system) located on the lid of a sample container, whence the GA reads information. Spectrometric analysis is performed automatically after the counting. Current status of GA can be remotely monitored. Information about the activities of the GA, measurement completion or failures of the equipment are automatically generated and sent to a mobile phone and the operator PC. A presentation of the GA is available at https://youtu.be/1lQhfo0Fljo. (orig.)

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

    International Nuclear Information System (INIS)

    Osanai, Masao

    1978-01-01

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

  8. Characterization of a lead breast shielding for dose reduction in computed tomography

    Energy Technology Data Exchange (ETDEWEB)

    Correia, Paula Duarte; Brochi, Marco Aurelio Corte; Azevedo-Marques, Paulo Mazzoncini de, E-mail: pauladuarte@usp.br [Universidade de Sao Paulo (FM/RSP), Ribeirao Preto, SP (Brazil). Faculdade de Medicina; Granzotti, Cristiano Roberto Fabri; Santos, Yago da Silva [Universidade de Sao Paulo (FFCLRP/RSP), Ribeirao Preto, SP (Brazil). Faculdade de Filosofia, Ciencias e Letras

    2014-07-15

    Objective: several studies have been published regarding the use of bismuth shielding to protect the breast in computed tomography (CT) scans and, up to the writing of this article, only one publication about barium shielding was found. The present study was aimed at characterizing, for the first time, a lead breast shielding. Materials and methods: the percentage dose reduction and the influence of the shielding on quantitative imaging parameters were evaluated. Dose measurements were made on a CT equipment with the aid of specific phantoms and radiation detectors. A processing software assisted in the qualitative analysis evaluating variations in average CT number and noise on images. Results: the authors observed a reduction in entrance dose by 30% and in CTDIvol by 17%. In all measurements, in agreement with studies in the literature, the utilization of cotton fiber as spacer object reduced significantly the presence of artifacts on the images. All the measurements demonstrated increase in the average CT number and noise on the images with the presence of the shielding. Conclusion: as expected, the data observed with the use of lead shielding were of the same order as those found in the literature about bismuth shielding. (author)

  9. Method of constructing shielding wall

    International Nuclear Information System (INIS)

    Nagao, Tetsuya.

    1990-01-01

    For instance, surfaces of lead particles each formed into a sphere of about 0.5 to 0.3 mm grain size are coated with a coating material of a synthetic resin comprising a polymeric material such as teflon. Subsequently, the floated lead particle are kneaded with concrete materials and then poured into a molding die by way of a hose. After coagulation, the molding die is removed to complete shielding walls in which lead particles are scattered substantially at an equal distance. In this way, since the lead particles are mixed into the shielding walls, shielding effects can be improved by so much as the lead particles are mixed, thereby enabling to reduce the thickness of the shielding walls. Further, since the lead particles are coated with the coating material, the lead particles are insulated from the concrete materials, thereby enabling to prevent the corrosion of the lead particles. Furthermore, since the lead particles and the concrete materials can be transported with ease, operation labors can be reduced. (T.M.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. Shielding Effectiveness of a Thin Film Window

    National Research Council Canada - National Science Library

    Johnson, Eric

    1998-01-01

    .... The predicted shielding effectiveness was 29 dB based on theoretical calculations. The error analysis of the shielding effectiveness showed that this predicted value was within the measurement error...

  12. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

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

  13. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2002-04-01

    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)

  14. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2003-01-01

    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)

  15. Sensitive detectors in HPLC

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Detection of sample components in HPLC is difficult for many reasons; the key difficulty is the mobile phase which usually has properties similar to the solute. A variety of detectors have been developed for use in HPLC based on one of the above approaches; however, the search is still continuing for an ideal or universal detector. A universal detector should have the following characteristics: (1) responds to all solutes or has predictable specificity; (2) high detectability and the same predictable response; (3) fast response; (4) wide range of linearity; (5) unaffected by changes in temperature and mobile-phase flow; (6) responds independently of the mobile phase; (7) makes no contribution to extracolumn band broadening; (8) reliable and convenient to use; (9) nondestructive to the solute; (10) provides qualitative information on the detected peak. Unfortunately, no available HPLC detector possesses all these properties. 145 refs

  16. Shielding of the GERDA experiment against external gamma background

    International Nuclear Information System (INIS)

    Barabanov, I.; Bezrukov, L.; Demidova, E.; Gurentsov, V.; Kianovsky, S.; Knoepfle, K.T.; Kornouhkov, V.; Schwingenheuer, B.; Vasenko, A.

    2009-01-01

    The GERmanium Detector Array (GERDA) experiment will search for neutrinoless double beta decay of 76 Ge and is currently under construction at the INFN Laboratori Nazionali del Gran Sasso (LNGS) in Italy. The basic design of GERDA is the use of cryogenic liquid and water of high purity as a superior shield against the hitherto dominant background from external gamma radiation. In this paper we show by Monte Carlo simulations and analytical calculations how GERDA was designed to suppress this background at Q ββ ( 76 Ge)=2039keV to a level of about 10 -4 cts/(keVkgy).

  17. Gamma ray and neutron shielding properties of some concrete materials

    International Nuclear Information System (INIS)

    Yilmaz, E.; Baltas, H.; Kiris, E.; Ustabas, I.; Cevik, U.; El-Khayatt, A.M.

    2011-01-01

    Highlights: → This study sheds light on the shielding properties of gamma-rays and neutrons for some concrete samples. → The experimental mass attenuation coefficients values were compared with theoretical values obtained using WinXCom. → Moreover, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. → The NXcom program was employed to calculate the attenuation coefficients values of neutrons. → These values showed a change with energy and composition of the concrete samples. - Abstract: Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.

  18. Investigations on the Broadband Shielding Effectiveness of Metallized Glass Fiber

    National Research Council Canada - National Science Library

    Coburn, William

    1998-01-01

    ...) is an E-glass fiber metallized with Al and processed into a nonwoven mat. When formed into a mat, the MGFs lead to an effective sample conductivity, sigma eff, which is the parameter of interest for electromagnetic shielding in the RF region...

  19. Detector for failed fuel elements

    International Nuclear Information System (INIS)

    Ito, Masaru.

    1979-01-01

    Purpose: To provide automatic monitor for the separation or reactor water and sampling water, in a failed fuel element detector using a sipping chamber. Constitution: A positional detector for the exact mounting of a sipping chamber on a channel box and a level detector for the detection of complete discharge of cooling water in the sipping chamber are provided in the sipping chamber. The positional detector is contacted to the upper end of the channel box and operated when the sipping chamber is correctly mounted to the fuel assemblies. The level detector comprises a float and a limit switch and it is operated when the water in the sipping chamber is discharged by a predetermined amount. Isolation of reactor water and sampling water are automatically monitored by the signal from these two detectors. (Ikeda, J.)

  20. Pretinning Nickel-Plated Wire Shields

    Science.gov (United States)

    Igawa, J. A.

    1985-01-01

    Nickel-plated copper shielding for wires pretinned for subsequent soldering with help of activated rosin flux. Shield cut at point 0.25 to 0.375 in. (6 to 10 mm) from cut end of outer jacket. Loosened end of shield straightened and pulled toward cut end. Insulation of inner wires kept intact during pretinning.

  1. Computed tomography shielding methods: a literature review.

    Science.gov (United States)

    Curtis, Jessica Ryann

    2010-01-01

    To investigate available shielding methods in an effort to further awareness and understanding of existing preventive measures related to patient exposure in computed tomography (CT) scanning. Searches were conducted to locate literature discussing the effectiveness of commercially available shields. Literature containing information regarding breast, gonad, eye and thyroid shielding was identified. Because of rapidly advancing technology, the selection of articles was limited to those published within the past 5 years. The selected studies were examined using the following topics as guidelines: the effectiveness of the shield (percentage of dose reduction), the shield's effect on image quality, arguments for or against its use (including practicality) and overall recommendation for its use in clinical practice. Only a limited number of studies have been performed on the use of shields for the eyes, thyroid and gonads, but the evidence shows an overall benefit to their use. Breast shielding has been the most studied shielding method, with consistent agreement throughout the literature on its effectiveness at reducing radiation dose. The effect of shielding on image quality was not remarkable in a majority of studies. Although it is noted that more studies need to be conducted regarding the impact on image quality, the currently published literature stresses the importance of shielding in reducing dose. Commercially available shields for the breast, thyroid, eyes and gonads should be implemented in clinical practice. Further research is needed to ascertain the prevalence of shielding in the clinical setting.

  2. Flame detector operable in presence of proton radiation

    Science.gov (United States)

    Walker, D. J.; Turnage, J. E.; Linford, R. M. F.; Cornish, S. D. (Inventor)

    1974-01-01

    A detector of ultraviolet radiation for operation in a space vehicle which orbits through high intensity radiation areas is described. Two identical ultraviolet sensor tubes are mounted within a shield which limits to acceptable levels the amount of proton radiation reaching the sensor tubes. The shield has an opening which permits ultraviolet radiation to reach one of the sensing tubes. The shield keeps ultraviolet radiation from reaching the other sensor tube, designated the reference tube. The circuitry of the detector subtracts the output of the reference tube from the output of the sensing tube, and any portion of the output of the sensing tube which is due to proton radiation is offset by the output of the reference tube. A delay circuit in the detector prevents false alarms by keeping statistical variations in the proton radiation sensed by the two sensor tubes from developing an output signal.

  3. High sensitivity isotope analysis with a 252Cf--235U fueled subcritical multiplier and low background photon detector systems

    International Nuclear Information System (INIS)

    Wogman, N.A.; Rieck, H.G. Jr.; Laul, J.C.; MacMurdo, K.W.

    1976-09-01

    A 252 Cf activation analysis facility has been developed for routine multielement analysis of a wide variety of solid and liquid samples. The facility contains six sources of 252 Cf totaling slightly over 100 mg. These sources are placed in a 93 percent 235 U-enriched uranium core which is subcritical with a K effective of 0.985 (multiplication factor of 66). The system produces a thermal flux on the order of 10 +1 neutrons per square centimeter per second. A pneumatic rabbit system permits automatic irradiation, decay, and counting regimes to be performed unattended on the samples. The activated isotopes are analyzed through their photon emissions with state-of-the-art intrinsic Ge detectors, Ge(Li) detectors, and NaI(Tl) multidimensional gamma ray spectrometers. High efficiency (25 percent), low background, anticoincidence shielded Ge(Li) gamma ray detector systems have been constructed to provide the lowest possible background, yet maintain a peak to Compton ratio of greater than 1000 to 1. The multidimensional gamma ray spectrometer systems are composed of 23 cm diameter x 20 cm thick NaI(Tl) crystals surrounded by NaI(Tl) anticoincidence shields. The detection limits for over 65 elements have been determined for this system. Over 40 elements are detectable at the 1 part per million level at a precision of +-10 percent

  4. BRH Gonad Shielding Program: where it has led

    International Nuclear Information System (INIS)

    Arcarese, J.S.

    1975-01-01

    Some topics discussed are: Bureau of Radiological Health guidelines; types of gonad shields; specific area shielding; gonad shielding guidelines; and publication of pamphlet on types of shields and circumstances under which they should be used

  5. Preparation of the in-house neutron detectors and the software needed to process experimental data

    International Nuclear Information System (INIS)

    Haddad, Kh.; Haj-Hassan, H.; Helal, W.

    2007-04-01

    In - house neutron activation detectors were prepared in this work using pure commercial gold. The neutron self-shielding factors in the foils for both thermal and epithermal neutrons have been determined experimentally. The work shows good results repeatability and good agreement with certified activation monitors. the software KHW for neutron flux measurements using local and standards gold foils was designed and performed locally. it deals as well with irradiated uranium spectrums to calculate some important fission product ratios for neutron flux measurement. Some experiments were performed to investigate the possibility of using uranium, produced in the pilot plant, as fission neutron detector. The results shows the possibility of using fission product ratios to determine the cooling time of the samples. It shows also the possibility of using fission and activation product ratios as an indicators of neutron fluences ratios.(author)

  6. Post Three Mile Island shielding review - a case history

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  7. Cryogenic detectors

    International Nuclear Information System (INIS)

    Zehnder, A.

    1987-01-01

    Presently the development of new large scale detector systems, used in very high energy physics experiments, is very active. In the low energy range, the introduction of charge coupled devices allows improved spacial and energy resolution. In the keV region, high resolution can only be achieved via the well established diffraction spectrometers with the well-known disadvantage of a small throughput. There exist no efficient detectors for non-ionizing radiation such as coherent nuclear scattering of weakly interacting particles. The development of high resolution solid state detectors in the keV-region with the possibility of nuclear recoil detection is therefore highly desired. Such detectors applied in astro and particle physics would thus allow one to obtain new information not achievable otherwise. Three types of cryogenic detectors exist: Calorimeters/Bolometers. This type is sensitive to the produced excess phonons and measures the deposited energy by detecting the heat pulses. Excess charge carriers should be used to produce phonons. Tunneling junctions. This type is sensitive to excess charge produced by the Cooper pair breakup. Excess phonons should be used to break up Cooper pairs. Superheated superconducting granules (SSG). An SSG detector consists of granules, the metastability of which is disturbed by radiation. The Meissner effect then causes a change in the field distribution of the applied external field, which can be detected. The present paper discusses the basic principle of calorimetric and tunneling junction detectors and some of their applications. 26 refs., 7 figs., 1 tab

  8. Evaluation of the performance of peridotite aggregates for radiation shielding concrete

    International Nuclear Information System (INIS)

    Wang, Jinjun; Li, Guofeng; Meng, Dechuan

    2014-01-01

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

  9. Detectors - Electronics

    International Nuclear Information System (INIS)

    Bregeault, J.; Gabriel, J.L.; Hierle, G.; Lebotlan, P.; Leconte, A.; Lelandais, J.; Mosrin, P.; Munsch, P.; Saur, H.; Tillier, J.

    1998-01-01

    The reports presents the main results obtained in the fields of radiation detectors and associated electronics. In the domain of X-ray gas detectors for the keV range efforts were undertaken to rise the detector efficiency. Multiple gap parallel plate chambers of different types as well as different types of X → e - converters were tested to improve the efficiency (values of 2.4% at 60 KeV were reached). In the field of scintillators a study of new crystals has been carried out (among which Lutetium orthosilicate). CdTe diode strips for obtaining X-ray imaging were studied. The complete study of a linear array of 8 CdTe pixels has been performed and certified. The results are encouraging and point to this method as a satisfying solution. Also, a large dimension programmable chamber was used to study the influence of temperature on the inorganic scintillators in an interval from -40 deg. C to +150 deg. C. Temperature effects on other detectors and electronic circuits were also investigated. In the report mentioned is also the work carried out for the realization of the DEMON neutron multidetector. For neutron halo experiments different large area Si detectors associated with solid and gas position detectors were realized. In the frame of a contract with COGEMA a systematic study of Li doped glasses was undertaken aiming at replacing with a neutron probe the 3 He counters presently utilized in pollution monitoring. An industrial prototype has been realised. Other studies were related to integrated analog chains, materials for Cherenkov detectors, scintillation probes for experiments on fundamental processes, gas position sensitive detectors, etc. In the field of associated electronics there are mentioned the works related to the multidetector INDRA, data acquisition, software gamma spectrometry, automatic gas pressure regulation in detectors, etc

  10. Survivor shielding. Part A. Nagasaki factory worker shielding

    International Nuclear Information System (INIS)

    Santoro, Robert T.; Barnes, John M.; Azmy, Yousry Y.; Kerr, George D.; Egbert, Stephen D.; Cullings, Harry M.

    2005-01-01

    Recent investigations based on conventional chromosome aberration data by the RERF suggest that the DS86 doses received by many Nagasaki factory workers may have been overestimated by as much as 40% relative to those for other survivors in Japanese-type houses and other shielding configurations (Kodama et al. 2001). Since the factory workers represent about 25% of the Nagasaki survivors with DS86 doses in excess of 0.5 Gy (50 rad), systematic errors in their dose estimates can have a major impact on the risk coefficients from RERF studies. The factory worker doses may have been overestimated for a number of reasons. The calculation techniques, including the factory building modeling, weapon source spectra and cross-section data used in the DS86 shielding calculations were not detailed enough to replicate actual conditions. The models used did not take into account local shielding provided by machinery, tools, and the internal structure in the buildings. In addition, changes in the disposition of shielding following collapse of the building by the blast wave were not considered. The location of large factory complexes may be uncertain, causing large numbers of factory survivors, correctly located relative to each other, to be uniformly too close to the hypocenter. Any or all of these reasons are sufficient to result in an overestimate of the factory worker doses. During the DS02 studies, factory worker doses have been reassessed by more carefully modeling the factory buildings, incorporating improved radiation transport methods and cross-section data and using the most recent bomb leakage spectra (Chapter 2). Two-dimensional discrete ordinates calculations were carried out initially to estimate the effects of workbenches and tools on worker doses to determine if the inclusion of these components would, in fact, reduce the dose by amounts consistent with the RERF observations (Kodama et al. 2001). (author)

  11. Calculation analysis of the thickness of radiation shield for the RIA equipment IP10

    International Nuclear Information System (INIS)

    Benar Bukit; Kristiyanti; Hari Nurcahyadi

    2011-01-01

    Calculation Analysis has been performed on the thickness of radiation shield for the design of the Radioimmunoassay (RIA) IP10 counters using five detectors arranged in parallel. The calculation is intended to ensure that the radiation on each detector does not influence each other. The radiation shield is made of lead. The calculation of lead thickness was based on the principle of the lead plates absorptive power toward the gamma ray of a certain energy. which is the function of linear absorption coefficient and the material thickness. Assuming the use of Iodium-125(I-125) source with an activity 10 µCi, and expecting an absorptive power of 95%, calculations showed that the required lead thickness is equal to 0,013 cm. Since lead is soft and its availability in the market is limited, lead plate of 2 mm thickness are used instead, so that counting result for the detectors do not influence each other. (author)

  12. Note: A 102 dB dynamic-range charge-sampling readout for ionizing particle/radiation detectors based on an application-specific integrated circuit (ASIC)

    Science.gov (United States)

    Pullia, A.; Zocca, F.; Capra, S.

    2018-02-01

    An original technique for the measurement of charge signals from ionizing particle/radiation detectors has been implemented in an application-specific integrated circuit form. The device performs linear measurements of the charge both within and beyond its output voltage swing. The device features an unprecedented spectroscopic dynamic range of 102 dB and is suitable for high-resolution ion and X-γ ray spectroscopy. We believe that this approach may change a widespread paradigm according to which no high-resolution spectroscopy is possible when working close to or beyond the limit of the preamplifier's output voltage swing.

  13. Hydrogen detector

    International Nuclear Information System (INIS)

    Kumagaya, Hiromichi; Yoshida, Kazuo; Sanada, Kazuo; Chigira, Sadao.

    1994-01-01

    The present invention concerns a hydrogen detector for detecting water-sodium reaction. The hydrogen detector comprises a sensor portion having coiled optical fibers and detects hydrogen on the basis of the increase of light transmission loss upon hydrogen absorption. In the hydrogen detector, optical fibers are wound around and welded to the outer circumference of a quartz rod, as well as the thickness of the clad layer of the optical fiber is reduced by etching. With such procedures, size of the hydrogen detecting sensor portion can be decreased easily. Further, since it can be used at high temperature, diffusion rate is improved to shorten the detection time. (N.H.)

  14. MUON DETECTORS: DT

    CERN Multimedia

    Marco Dallavalle

    2013-01-01

    The DT group is undertaking substantial work both for detector maintenance and for detec-tor upgrade. Maintenance interventions on chambers and minicrates require close collaboration between DT, RPC and HO, and are difficult because they depend on the removal of thermal shields and cables on the front and rear of the chambers in order to gain access. The tasks are particularly critical on the central wheel due to the presence of fixed services. Several interventions on the chambers require extraction of the DT+RPC package: a delicate operation due to the very limited space for handling the big chambers, and the most dangerous part of the DT maintenance campaign. The interventions started in July 2013 and will go on until spring 2014. So far out of the 16 chambers with HV problems, 13 have been already repaired, with a global yield of 217 recovered channels. Most of the observed problems were due to displacement of impurities inside the gaseous volume. For the minicrates and FE, repairs occurred on 22 chambe...

  15. System for imaging plutonium through heavy shielding

    International Nuclear Information System (INIS)

    Kuckertz, T.H.; Cannon, T.M.; Fenimore, E.E.; Moss, C.E.; Nixon, K.V.

    1984-04-01

    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 57 Co unshielded at a distance of 27 m and 220 g of bulk plutonium (11.8% 240 Pu) 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

  16. Mechanical properties of JPDR biological shield concrete

    International Nuclear Information System (INIS)

    Idei, Yoshio; Kamata, Hiroshi; Akutsu, Youichi; Onizawa, Kunio; Nakajima, Nobuya; Sukegawa, Takenori; Kakizaki, Masayoshi.

    1990-11-01

    Plant life of nuclear power plant will be determined by the aging degradation of main components and structures because of the difficulty and the cost of the replacement. These components are the reactor pressure vessel, concrete structures and cables. Authors have performed the investigation of JPDR biological shield which was the succeeded in first generating electricity in Japan and is now being decommissioned in JAERI. The test core samples were bored from the shield concrete and tested to obtain the mechanical properties. Test results are summarized as below, (1) Peak value of fast neutron dose was estimated as 1 x 10 18 n/cm 2 which is equivalent to the dose at the end of life for commercial power reactor. (2) Averaged compressive strength of all specimens had been increased about 20 % compared with initial design strength. (3) It was identified that the compressive strength had a little trend to increase with the increase of neutron dose within the dose range obtained in this study. (4) Tensile strength, Elastic modulus and Poisson's ratio showed little effect of neutron dose. (5) It was suggested that the inside and the mid-section liners were effective to keep the water in concrete and to avoid the reduction in strength. (author)

  17. DUMAND detector

    CERN Multimedia

    This object is one of the 256 other detectors of the DUMAND (Deep Underwater Muon And Neutrino Detection) experiment. The goal of the experiment was the construction of the first deep ocean high energy neutrino detector, to be placed at 4800 m depth in the Pacific Ocean off Keahole Point on the Big Island of Hawaii. A few years ago, a European conference with Cosmic experiments was organized at CERN as they were projects like DUMAND in Hawaii. Along with the conference, a temporary exhibition was organised as well. It was a collaboration of institutions from Germany, Japan, Switzerland and the U.S.A. CERN had borrowed equipment and objects from different institutes around the world, including this detector of the DUMAND experiment. Most of the equipment were sent back to the institutes, however this detector sphere was offered to a CERN member of the personnel.

  18. Detector applications

    International Nuclear Information System (INIS)

    Pehl, R.H.

    1977-10-01

    Semiconductor detectors are now applied to a very wide range of problems. The combination of relatively low cost, excellent energy resolution, and simultaneous broad energy-spectrum analysis is uniquely suited to many applications in both basic and applied physics. Alternative techniques, such as magnetic spectrometers for charged-particle spectroscopy, while offering better energy resolution, are bulky, expensive, and usually far more difficult to use. Furthermore, they do not directly provide the broad energy-spectrum measurements easily accomplished using semiconductor detectors. Scintillation detectors, which are approximately equivalent to semiconductor detectors in convenience and cost, exhibit 10 to 100 times worse energy resolution. However, their high efficiency and large potential size recommend their use in some measurements

  19. New facility shield design criteria

    International Nuclear Information System (INIS)

    Howell, W.P.

    1981-07-01

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

  20. INERT GAS SHIELD FOR WELDING

    Science.gov (United States)

    Jones, S.O.; Daly, F.V.

    1958-10-14

    S>An inert gas shield is presented for arc-welding materials such as zirconium that tend to oxidize rapidly in air. The device comprises a rectangular metal box into which the welding electrode is introduced through a rubber diaphragm to provide flexibility. The front of the box is provided with a wlndow having a small hole through which flller metal is introduced. The box is supplied with an inert gas to exclude the atmosphere, and with cooling water to promote the solidification of the weld while in tbe inert atmosphere. A separate water-cooled copper backing bar is provided underneath the joint to be welded to contain the melt-through at the root of the joint, shielding the root of the joint with its own supply of inert gas and cooling the deposited weld metal. This device facilitates the welding of large workpieces of zirconium frequently encountered in reactor construction.

  1. 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.

  2. Transient shielded liquid hydrogen containers

    International Nuclear Information System (INIS)

    Varghese, A.P.; Herring, R.H.

    1990-01-01

    The storage of hydrogen in the liquid phase has been limited in duration due to the thermal performance constraints of conventional Liquid Hydrogen containers available. Conventional Liquid Hydrogen containers lose hydrogen because of their relatively high heat leak and variations in usage pattern of hydrogen due to shutdowns. Local regulations also discourage venting of hydrogen. Long term storage of Liquid Hydrogen without product loss was usually accomplished using Liquid Nitrogen sacrificial shields. This paper reports on a new low heat leak container developed and patented that will extend the storage time of liquid hydrogen by five hundred percent. The principle of operation of the Transient Shields which makes the extraordinary performance of this container feasible is described in this paper. Also covered are the impact of this new container on present applications of hydrogen and the new opportunities afforded to Liquid hydrogen in the world hydrogen market

  3. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

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

    1975-08-01

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

  4. Radiation-shielding transparent material

    International Nuclear Information System (INIS)

    Kusumeki, Asao.

    1983-01-01

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

  5. Analysis of portable gamma flaw detectors concerning radiation hygiene

    International Nuclear Information System (INIS)

    Makarova, T.V.

    1982-01-01

    Design and shields of gamma flaw detectors as one of the main factors responsible for personnel dose were studied. The analysis was conducted using the results of radiation hygienic surveys of gamma flaw detection laboratories functioning constantly in Estonia. It is shown that recently the replacement of GUP apparatuses by flaw detectors of RID and ''Gamma-RID'' (types which have design and shielding advantages is observed. However personnel doses have not reduced considerably for the last 10 years. This fact is attributed to design disadvantages of the RID and ''Gamma-RID'' apparatuses the removing of which will give the decreasing of annual personnel dose by 80 %

  6. Smoke detectors

    International Nuclear Information System (INIS)

    Bryant, J.; Howes, J.H.; Smout, D.W.S.

    1979-01-01

    A smoke detector is described which provides a smoke sensing detector and an indicating device and in which a radioactive substance is used in conjunction with two ionisation chambers. The system includes an outer electrode, a collector electrode and an inner electrode which is made of or supports the radioactive substance which, in this case, is 241 Am. The invention takes advantage of the fact that smoke particles can be allowed to enter freely the inner ionisation chamber. (U.K.)

  7. Radiation detector

    International Nuclear Information System (INIS)

    Gillies, W.

    1980-01-01

    The radiation detector for measuring e.g. a neutron flux consists of a central emitter, an insulating shell arranged around it, and a tube-shaped collector enclosing both. The emitter itself is composed of a great number of stranded, spiral wires of small diameter giving a defined flexibility to the detector. For emitter material Pt, Rh, V, Co, Ce, Os or Ta may be used. (DG) [de

  8. Radiation shield vest and skirt

    International Nuclear Information System (INIS)

    Maine, G.J.

    1982-01-01

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

  9. Handbook of radiation shielding data

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1976-07-01

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

  10. Shielding wall for thermonuclear device

    International Nuclear Information System (INIS)

    Uchida, Takaho.

    1989-01-01

    This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

  11. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    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)

  12. 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

  13. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Realization of the electrical Sentinel 4 detector integration

    Science.gov (United States)

    Hermsen, M.; Hohn, R.; Skegg, M.; Woffinden, C.; Reulke, R.

    2017-09-01

    The detectors of the Sentinel 4 multi spectral imager are operated in flight at 215K while the analog electronics is operated at ambient temperature. The detector is cooled by means of a radiator. For thermal reasons no active component has been allowed in the cooled area closest to the detector as the passive radiator is restricted in its size. For thermal decoupling of detector and electronics a long distance between detector and electronics is considered ideal as thermal conductivity decreases with the length of the connection. In contradiction a short connection between detector and electronics is ideal for the electronic signals. Only a short connection ensures the signal integrity of both the weak detector output signal but similarly also the clock signals for driving the detector. From a mechanical and thermal point of view the connection requires a certain minimum length. The selected solution serves all these needs but had to approach the limits of what is electrically, mechanically and thermally feasible. In addition, shielding from internal (self distortion) and external distorting signals has to be realized for the connection between FEE(Front End Electronics) and detectors. At the time of the design of the flex it was not defined whether the mechanical structure between FEE and FPA (Focal Plane Assembly) would act as a shielding structure. The physical separation between CCD detector and the Front-end Electronics, the adverse EMI environment in which the instrument will be operated in (the location of the instrument on the satellite is in vicinity to a down-link K-band communication antenna of the S/C) require at least the video output signals to be shielded. Both detectors (a NIR and a UVVIS detector) are sensitive to contamination and difficult to be cleaned in case of any contamination. This brings up extreme cleanliness requirements for the detector in manufacturing and assembly. Effectively the detector has to be kept in an ISO 5 environment and

  15. The shielding performance of multilayer composite shielding structures to 14.8 MeV fast neutrons

    International Nuclear Information System (INIS)

    Shen Zhiqiang; Kang Qing; Xu Jun; Wang Zhenggang; Lu Nan

    2014-01-01

    Cement-based round thin-layer samples mixed with 30% quality content of barite, and 20% quality content of carbide boron has Prepared, the same-diameter sliced samples of pure graphite and pure polyethylene has cut, then, samples combination and cross stack order has designed, formed four species Multilayer Composite shield structure, at last, neutron attenuation measurements has been done by experimental system of using 14.8 MeV neutrons from the 5SDH-2 accelerator and long counter composition, penetrating rate of samples and the shield structure to 14.8 MeV fast neutron has tested, and attenuation section has calculated. Results show that 14.8 MeV fast neutrons to higher penetration rates of thin layer samples, attenuation cross section of samples distinguish small between each other, must be increasing the thickness of the samples to reduce the experimental uncertainty; through composed of attenuation cross section and thickness parameters of composite structure, can more accurately predict the shielding ability of composite structures, error between calculation results and experimental results in 4%. (authors)

  16. Developing fine-pixel CdTe detectors for the next generation of high-resolution hard x-ray telescopes

    Science.gov (United States)

    Christe, Steven

    Over the past decade, the NASA Marshall Space Flight Center (MSFC) has been improving the angular resolution of hard X-ray (HXR; 20 "70 keV) optics to the point that we now routinely manufacture optics modules with an angular resolution of 20 arcsec Half Power Diameter (HDP), almost three times the performance of NuSTAR optics (Ramsey et al. 2013; Gubarev et al. 2013a; Atkins et al. 2013). New techniques are currently being developed to provide even higher angular resolution. High angular resolution HXR optics require detectors with a large number of fine pixels in order to adequately sample the telescope point spread function (PSF) over the entire field of view. Excessively over-sampling the PSF will increase readout noise and require more processing with no appreciable increase in image quality. An appropriate level of over-sampling is to have 3 pixels within the HPD. For the HERO mirrors, where the HPD is 26 arcsec over a 6-m focal length converts to 750 μm, the optimum pixel size is around 250 μm. At a 10-m focal length these detectors can support a 16 arcsec HPD. Of course, the detectors must also have high efficiency in the HXR region, good energy resolution, low background, low power requirements, and low sensitivity to radiation damage (Ramsey 2001). The ability to handle high counting rates is also desirable for efficient calibration. A collaboration between Goddard Space Flight Center (GSFC), MSFC, and Rutherford Appleton Laboratory (RAL) in the UK is developing precisely such detectors under an ongoing, funded APRA program (FY2015 to FY2017). The detectors use the RALdeveloped Application Specific Integrated Circuit (ASIC) dubbed HEXITEC, for High Energy X-Ray Imaging Technology. These HEXITEC ASICs can be bonded to 1- or 2- mm-thick Cadmium Telluride (CdTe) or Cadmium-Zinc-Telluride (CZT) to create a fine (250 μm pitch) HXR detector (Jones et al. 2009; Seller et al. 2011). The objectives of this funded effort are to develop and test a HEXITEC

  17. Mixture formation of direct gasoline injection engine. In cylinder gas sampling using fast response ionization detector; Tonai funsha gasoline engine no kongoki keisei. Kosoku FID ni yoru tonai gas sampling

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, H; Marubara, M; Ota, N; Kudo, H; Yamamoto, H [Mazda Motor Corp., Hiroshima (Japan)

    1997-10-01

    Local mixture concentration near the spark plug of a direct gasoline injection engine was observed by a fast flame ionization detector. To ensure combustion stability and good fuel economy in DISC operation, the swirl ratio and the piston configuration were optimized. Swirl is needed to retain well-vaporized and stable mixture near the spark plug especially in light load. And adequate volume in piston cavity is required for trapping curved fuel spray in it. With these specifications, the fuel economy improvement of 13 to 30 % was realized. 2 refs., 13 figs., 1 tab.

  18. Comparative study of radiation shielding parameters for bismuth borate glasses

    International Nuclear Information System (INIS)

    Kaundal, Rajinder Singh

    2016-01-01

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi 2 O 3- (1-x) B 2 O 3 where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of the prepared glass samples. (author)

  19. Comparative study of radiation shielding parameters for bismuth borate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Kaundal, Rajinder Singh, E-mail: rajinder_apd@yahoo.com [Department of Physics, School of Physical Sciences, Lovely Professional University, Phagwara, Punjab (India)

    2016-07-15

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi{sub 2}O{sub 3-}(1-x) B{sub 2}O{sub 3} where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of the prepared glass samples. (author)

  20. Neutron detector development at Brookhaven

    International Nuclear Information System (INIS)

    Yu, B.; Harder, J.A.; Mead, J.A.; Radeka, V.; Schaknowski, N.A.; Smith, G.C.

    2003-01-01

    Two-dimensional thermal neutron detectors have been the subject of research and development at Brookhaven for over 20 years. Based primarily on multi-wire chambers filled with a gas mixture containing 3 He, these detectors have been used in wide-ranging studies of molecular biology and material science samples. At each phase of development, experimenters have sought improvements in key parameters such as position resolution, counting rate, efficiency, solid-angle coverage and stability. A suite of detectors has been developed with sensitive areas ranging from 5x5 to 50x50 cm 2 . These devices incorporate low-noise-position readout and the best position resolution for thermal neutron gas detectors. Recent developments include a 1.5 mx20 cm detector containing multiple segments with continuously sensitive readout, and detectors with unity gain for ultra-high rate capability and long-term stability

  1. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Muon detector for the COSINE-100 experiment

    Science.gov (United States)

    Prihtiadi, H.; Adhikari, G.; Adhikari, P.; Barbosa de Souza, E.; Carlin, N.; Choi, S.; Choi, W. Q.; Djamal, M.; Ezeribe, A. C.; Ha, C.; Hahn, I. S.; Hubbard, A. J. F.; Jeon, E. J.; Jo, J. H.; Joo, H. W.; Kang, W.; Kang, W. G.; Kauer, M.; Kim, B. H.; Kim, H.; Kim, H. J.; Kim, K. W.; Kim, N. Y.; Kim, S. K.; Kim, Y. D.; Kim, Y. H.; Kudryavtsev, V. A.; Lee, H. S.; Lee, J.; Lee, J. Y.; Lee, M. H.; Leonard, D. S.; Lim, K. E.; Lynch, W. A.; Maruyama, R. H.; Mouton, F.; Olsen, S. L.; Park, H. K.; Park, H. S.; Park, J. S.; Park, K. S.; Pettus, W.; Pierpoint, Z. P.; Ra, S.; Rogers, F. R.; Rott, C.; Scarff, A.; Spooner, N. J. C.; Thompson, W. G.; Yang, L.; Yong, S. H.

    2018-02-01

    The COSINE-100 dark matter search experiment has started taking physics data with the goal of performing an independent measurement of the annual modulation signal observed by DAMA/LIBRA. A muon detector was constructed by using plastic scintillator panels in the outermost layer of the shield surrounding the COSINE-100 detector. It detects cosmic ray muons in order to understand the impact of the muon annual modulation on dark matter analysis. Assembly and initial performance tests of each module have been performed at a ground laboratory. The installation of the detector in the Yangyang Underground Laboratory (Y2L) was completed in the summer of 2016. Using three months of data, the muon underground flux was measured to be 328 ± 1(stat.)± 10(syst.) muons/m2/day. In this report, the assembly of the muon detector and the results from the analysis are presented.

  4. A mower detector to judge soil sorting

    International Nuclear Information System (INIS)

    Bramlitt, E.T.; Johnson, N.R.

    1995-01-01

    Thermo Nuclear Services (TNS) has developed a mower detector as an inexpensive and fast means for deciding potential value of soil sorting for cleanup. It is a shielded detector box on wheels pushed over the ground (as a person mows grass) at 30 ft/min with gamma-ray counts recorded every 0.25 sec. It mirror images detection by the TNS transportable sorter system which conveys soil at 30 ft/min and toggles a gate to send soil on separate paths based on counts. The mower detector shows if contamination is variable and suitable for sorting, and by unique calibration sources, it indicates detection sensitivity. The mower detector has been used to characterize some soil at Department of Energy sites in New Jersey and South Carolina

  5. A mower detector to judge soil sorting

    Energy Technology Data Exchange (ETDEWEB)

    Bramlitt, E.T.; Johnson, N.R. [Thermo Nuclear Services, Inc., Albuquerque, NM (United States)

    1995-12-31

    Thermo Nuclear Services (TNS) has developed a mower detector as an inexpensive and fast means for deciding potential value of soil sorting for cleanup. It is a shielded detector box on wheels pushed over the ground (as a person mows grass) at 30 ft/min with gamma-ray counts recorded every 0.25 sec. It mirror images detection by the TNS transportable sorter system which conveys soil at 30 ft/min and toggles a gate to send soil on separate paths based on counts. The mower detector shows if contamination is variable and suitable for sorting, and by unique calibration sources, it indicates detection sensitivity. The mower detector has been used to characterize some soil at Department of Energy sites in New Jersey and South Carolina.

  6. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  7. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Nonlinear AC susceptibility, surface and bulk shielding

    Science.gov (United States)

    van der Beek, C. J.; Indenbom, M. V.; D'Anna, G.; Benoit, W.

    1996-02-01

    We calculate the nonlinear AC response of a thin superconducting strip in perpendicular field, shielded by an edge current due to the geometrical barrier. A comparison with the results for infinite samples in parallel field, screened by a surface barrier, and with those for screening by a bulk current in the critical state, shows that the AC response due to a barrier has general features that are independent of geometry, and that are significantly different from those for screening by a bulk current in the critical state. By consequence, the nonlinear (global) AC susceptibility can be used to determine the origin of magnetic irreversibility. A comparison with experiments on a Bi 2Sr 2CaCu 2O 8+δ crystal shows that in this material, the low-frequency AC screening at high temperature is mainly due to the screening by an edge current, and that this is the unique source of the nonlinear magnetic response at temperatures above 40 K.

  9. Lunar soil as shielding against space radiation

    Energy Technology Data Exchange (ETDEWEB)

    Miller, J. [Lawrence Berkeley National Laboratory, MS 83R0101, 1 Cyclotron Road, Berkeley, CA 94720 (United States)], E-mail: miller@lbl.gov; Taylor, L. [Planetary Geosciences Institute, Department of Earth and Planetary Sciences, University of Tennessee, Knoxville, TN 37996 (United States); Zeitlin, C. [Southwest Research Institute, Boulder, CO 80302 (United States); Heilbronn, L. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Guetersloh, S. [Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); DiGiuseppe, M. [Northrop Grumman Corporation, Bethpage, NY 11714 (United States); Iwata, Y.; Murakami, T. [National Institute of Radiological Sciences, Chiba 263-8555 (Japan)

    2009-02-15

    We have measured the radiation transport and dose reduction properties of lunar soil with respect to selected heavy ion beams with charges and energies comparable to some components of the galactic cosmic radiation (GCR), using soil samples returned by the Apollo missions and several types of synthetic soil glasses and lunar soil simulants. The suitability for shielding studies of synthetic soil and soil simulants as surrogates for lunar soil was established, and the energy deposition as a function of depth for a particular heavy ion beam passing through a new type of lunar highland simulant was measured. A fragmentation and energy loss model was used to extend the results over a range of heavy ion charges and energies, including protons at solar particle event (SPE) energies. The measurements and model calculations indicate that a modest amount of lunar soil affords substantial protection against primary GCR nuclei and SPE, with only modest residual dose from surviving charged fragments of the heavy beams.

  10. Shaped detector

    International Nuclear Information System (INIS)

    Carlson, R.W.

    1981-01-01

    A radiation detector or detector array which has a non-constant spatial response, is disclosed individually and in combination with a tomographic scanner. The detector has a first dimension which is oriented parallel to the plane of the scan circle in the scanner. Along the first dimension, the detector is most responsive to radiation received along a centered segment of the dimension and less responsive to radiation received along edge segments. This non-constant spatial response can be achieved in a detector comprised of a scintillation crystal and a photoelectric transducer. The scintillation crystal in one embodiment is composed of three crystals arranged in layers, with the center crystal having the greatest light conversion efficiency. In another embodiment, the crystal is covered with a reflective substance around the center segment and a less reflective substance around the remainder. In another embodiment, an optical coupling which transmits light from adjacent the center segment with the greatest intensity couples the scintillation crystal and the photoelectric transducer. In yet another embodiment, the photoelectric transducer comprises three photodiodes, one receiving light produced adjacent the central segment and the other two receiving light produced adjacent the edge segments. The outputs of the three photodiodes are combined with a differential amplifier

  11. A neutron activation detector

    International Nuclear Information System (INIS)

    Ambardanishvili, T.S.; Kolomiitsev, M.A.; Zakharina, T.Y.; Dundua, V.J.; Chikhladze, N.V.

    1973-01-01

    The present invention concerns a neutron activation detector made from a moulded and hardened composition. According to the invention, that composition contains an activable substance constituted by at least two chemical elements and/or compounds of at least two chemical elements. Each of these chemical elements is capable of reacting with the neutrons forming radio-active isotopes with vatious levels of energy during desintegration. This neutron detector is mainly suitable for measuring integral thermal neutron and fast neutron fluxes during irradiation of the sample, and also for measuring the intensities of neutron fields [fr

  12. Passive detectors for neutron fluence measurement

    International Nuclear Information System (INIS)

    Holt, P.D.

    1985-01-01

    The use of neutron activation detectors (slow neutron detectors and threshold detectors) and fission track detectors for radiological protection purposes, principally in criticality dosimetry, dosimetry of pulsed accelerators and calibration of neutron fluxes is discussed. References are given to compilations of cross sections. For the determination of the activity induced, either beta ray or gamma ray counting may be used. For beta-ray counting, thin foils are usually necessary which result in low neutron sensitivity. When fission track detectors are used, it is necessary to know the efficiency of track registration. Alternatively, a detector-counter system may be calibrated by exposure to a known flux of monoenergetic neutrons. Usually, the sensitivity of activation detectors is low because small foils are used. For criticality dosimetry, calibration work and shielding studies on accelerators, low sensitivity is acceptable. However, there are some instances where, by the use of long integration times, or very large quantities of detector material with gamma ray detection, neutron fluences in operational areas have been measured. (author)

  13. Research on Primary Shielding Calculation Source Generation Codes

    Science.gov (United States)

    Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun

    2017-09-01

    Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.

  14. Locating gamma radiation source by self collimating BGO detector system

    Energy Technology Data Exchange (ETDEWEB)

    Orion, I; Pernick, A; Ilzycer, D; Zafrir, H [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center; Shani, G [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1996-12-01

    The need for airborne collimated gamma detector system to estimate the radiation released from a nuclear accident has been established. A BGO detector system has been developed as an array of separate seven cylindrical Bismuth Germanate scintillators, one central detector symmetrically surrounded by six detectors. In such an arrangement, each of the detectors reduced the exposure of other detectors in the array to a radiation incident from a possible specific spatial angle, around file array. This shielding property defined as `self-collimation`, differs the point source response function for each of the detectors. The BGO detector system has a high density and atomic number, and therefore provides efficient self-collimation. Using the response functions of the separate detectors enables locating point sources as well as the direction of a nuclear radioactive plume with satisfactory angular resolution, of about 10 degrees. The detector`s point source response, as function of the source direction, in a horizontal plane, has been predicted by analytical calculation, and was verified by Monte-Carlo simulation using the code EGS4. The detector`s response was tested in a laboratory-scale experiment for several gamma ray energies, and the experimental results validated the theoretical (analytical and Monte-Carlo) results. (authors).

  15. Pb-free Radiation Shielding Glass Using Coal Fly Ash

    Directory of Open Access Journals (Sweden)

    Watcharin Rachniyom

    2015-12-01

    Full Text Available In this work, Pb-free shielding glass samples were prepared by the melt quenching technique using subbituminous fly ash (SFA composed of xBi2O3 : (60-xB2O3 : 10Na2O : 30SFA (where x = 10, 15, 20, 25, 30 and 35 by wt%. The samples were investigated for their physical and radiation shielding properties. The density and hardness were measured. The results showed that the density increased with the increase of Bi2O3 content. The highest value of hardness was observed for glass sample with 30 wt% of Bi2O3 concentration. The samples were investigated under 662 keV gamma ray and the results were compared with theoretical calculations. The values of the mass attenuation coefficient (μm, the atomic cross section (σe and the effective atomic number (Zeff were found to increase with an increase of the Bi2O3 concentration and were in good agreement with the theoretical calculations. The best results for the half-value layer (HVL were observed in the sample with 35 wt% of Bi2O3 concentration, better than the values of barite concrete. These results demonstrate the viability of using coal fly ash waste for radiation shielding glass without PbO in the glass matrices.

  16. RID-41 gamma flaw detector

    International Nuclear Information System (INIS)

    Glebov, V.N.; Zubkov, V.S.; Majorov, A.N.; Murashev, A.I.; Firstov, V.G.; Yampol'skij, V.V.; Goncharov, V.I.; Sakhanov, A.S.

    1978-01-01

    The design is described and the main characteristics are given of a universal stationary hose-type gamma flow detector with a 60 Co source from 3O to 4g0 Ci for high-productive control of thick-walled products from steel and other materials. The principal units of the instrument are a radiation head, a control panel, and a charge-exchange container. The flaw detector may be used both in shield chambers and in shop or mounting conditions on complying with due requirements of radiation protection. The high activity of the source at relatively small dimensions of its active part ensures good detection of defects. The high radioscopy rate permits to use the flaw detector in conditions of increased background radiation, e.g. during routine repairs and inspections at nuclear power plants. The instrument may also be used in radiometric complexes, and produces a considerable economic effect. This flaw-detector corresponds to ISO and IAEA requirements and may be delivered for export

  17. Application of the self-powered detector concept in the design of a threshold gamma-ray detector

    International Nuclear Information System (INIS)

    LeVert, F.E.

    1979-01-01

    The self-powered detector concept has been utilized to develop an energy threshold gamma-ray detector. Gamma-ray energy discrimination is achieved by using a thick annular lead shield around the outer wall (emitter) of the detector in conjunction with a self-shielding central electrode (collector). Measurements conducted in the graphite pit of the Argonne Thermal Source Reactor have confirmed its ability to detect high-energy prompt fission gamma rays while discriminating against a significant flux of low-energy gamma rays from the decay of fission products. Also, auto-power spectral densities obtained with the detector were used to estimate the kinetic parameter, β/l, of the reactor

  18. Shielded room measurements, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Stanton, J.S.

    1949-02-22

    The attenuation of electro-statically and electro-magnetically shielded rooms in the ``E,`` ``R,`` ``I,`` and ``T`` Buildings was measured so that corrective measure could be taken if the attenuation was found to be low. If remedial measures could not be taken, the shortcomings of the rooms would be known. Also, the men making the measurements should oversee construction and correct errors at the time. The work was performed by measuring the attenuation at spot frequencies over the range of from 150 kilocycles to 1280 megacycles with suitable equipment mounted in small rubber-tried trucks. The attenuation was determined by ``before and after`` shielding and/or ``door open and door closed`` measurements after installation of copper shielding. In general, attenuation in the frequency range of approximately 10 to 150 mc. was good and was of the order expected. At frequencies in the range of 150 mc. to 1280 mc., the attenuation curve was more erratic; that is, at certain frequencies a severe loss of attenuation was noted, while at others, the attenuation was very good. This was mainly due to poor or faulty seals around doors and pass windows. These poor seals existed in the ``T,`` ``E,`` and ``I`` Buildings because the doors were fitted improperly and somewhat inferior material was used. By experience from these difficulties, both causes were corrected in the ``R`` Building, which resulted in the improvement of the very high frequency (v.h.f.) range in this building. In some specific cases, however, the results were about the same. For the range of frequencies below approximately 10 mc., the attenuation, in almost all cases, gradually decreased as the frequency decreased and reached a minimum at .3 to 1.0 mc. This loss of attenuation was attributed to multiple grounding caused by moisture in the insulating timbers and will gradually decrease as the wood dries out.

  19. Magnetic shielding of a limiter

    International Nuclear Information System (INIS)

    Brevnov, N.N.; Stepanov, S.B.; Khimchenko, L.N.; Matthews, G.F.; Goodal, D.H.J.

    1991-01-01

    Localization of plasma interaction with material surfaces in a separate chamber, from where the escape of impurities is hardly realized, i.e. application of magnetic divertors or pump limiters, is the main technique for reduction of the impurity content in a plasma. In this case, the production of a divertor configuration requires a considerable power consumption and results in a less effective utilization of the magnetic field volume. Utilization of a pump limiter, for example the ICL-type, under tokamak-reactor conditions would result in the extremely high and forbidden local heat loadings onto the limiter surface. Moreover, the magnetically-shielded pump limiter (MSL) was proposed to combine positive properties of the divertor and the pump limiter. The idea of magnetic shielding is to locate the winding with current inside the limiter head so that the field lines of the resultant magnetic field do not intercept the limiter surface. In this case the plasma flows around the limiter leading edges and penetrates into the space under the limiter. The shielding magnetic field can be directed either counter the toroidal field or counter the poloidal one of a tokamak, dependent on the concrete diagram of the device. Such a limiter has a number of advantages: -opportunity to control over the particle and impurity recycling without practical influence upon the plasma column geometry, - perturbation of a plasma column magnetic configuration from the side of such a limiter is less than that from the side of the divertor coils. The main deficiency is the necessity to locate active windings inside the discharge chamber. (author) 5 refs., 3 figs

  20. BES detector

    International Nuclear Information System (INIS)

    Bai, J.Z.; Bian, Q.; Chen, G.M.; Chen, L.J.; Chen, S.N.; Chen, Y.Q.; Chen, Z.Q.; Chi, Y.K.; Cui, H.C.; Cui, X.Z.; Deng, S.S.; Deng, Y.W.; Ding, H.L.; Dong, B.Z.; Dong, X.S.; Du, X.; Du, Z.Z.; Feng, C.; Feng, Z.; Fu, Z.S.; Gao, C.S.; Gao, M.L.; Gao, S.Q.; Gao, W.X.; Gao, Y.N.; Gu, S.D.; Gu, W.X.; Guan, Y.Z.; Guo, H.F.; Guo, Y.N.; Guo, Y.Y.; Han, S.W.; Han, Y.; Hao, W.; He, J.; He, K.R.; He, M.J.; Hou, X.J.; Hu, G.Y.; Hu, J.S.; Hu, J.W.; Huang, D.Q.; Huang, Y.Z.; Jia, Q.P.; Jiang, C.H.; Ju, Q.; Lai, Y.F.; Lang, P.F.; Li, D.S.; Li, F.; Li, H.; Li Jia; Li, J.T.; Li Jin; Li, L.L.; Li, P.Q.; Li, Q.M.; Li, R.B.; Li, S.Q.; Li, W.; Li, W.G.; Li, Z.X.; Liang, G.N.; Lin, F.C.; Lin, S.Z.; Lin, W.; Liu, Q.; Liu, R.G.; Liu, W.; Liu, X.; Liu, Z.A.; Liu, Z.Y.; Lu, C.G.; Lu, W.D.; Lu, Z.Y.; Lu, J.G.; Ma, D.H.; Ma, E.C.; Ma, J.M.; Mao, H.S.; Mao, Z.P.; Meng, X.C.; Ni, H.L.; Nie, J.; Nie, Z.D.; Niu, W.P.; Pan, L.J.; Qi, N.D.; Qian, J.J.; Qu, Y.H.; Que, Y.K.; Rong, G.; Ruan, T.Z.; Shao, Y.Y.; Shen, B.W.; Shen, D.L.; Shen, J.; Sheng, H.Y.; Sheng, J.P.; Shi, H.Z.; Song, X.F.; Sun, H.S.; Tang, F.K.; Tang, S.Q.; Tian, W.H.; Wang, F.; Wang, G.Y.; Wang, J.G.; Wang, J.Y.; Wang, L.S.; Wang, L.Z.; Wang, M.; Wang, P.; Wang, P.L.; Wang, S.M.; Wang, S.Q.; Wang, T.J.; Wang, X.W.; Wang, Y.Y.; Wang, Z.H.; Wang, Z.J.; Wei, C.L.; Wei, Z.Z.; Wu, J.W.; Wu, S.H.; Wu, S.Q.; Wu, W.M.; Wu, X.D.; Wu, Z.D.; Xi, D.M.; Xia, X.M.; Xiao, J.; Xie, P.P.; Xie, X.X.; Xu, J.G.; Xu, R.S.; Xu, Z.Q.; Xuan, B.C.; Xue, S.T.; Yan, J.; Yan, S.P.; Yan, W.G.; Yang, C.Z.; Yang, C.M.; Yang, C.Y.; Yang, X.F.; Yang, X.R.; Ye, M.H.; Yu, C.H.; Yu, C.S.; Yu, Z.Q.; Zhang, B.Y.; Zhang, C.D.; Zhang, C.C.; Zhang, C.Y.; Zhang, D.H.; Zhang, G.; Zhang, H.Y.; Zhang, H.L.; Zhang, J.W.; Zhang, L.S.; Zhang, S.Q.; Zhang, Y.P.; Zhang, Y.; Zhang, Y.M.; Zhao, D.X.; Zhao, J.W.; Zhao, M.; Zhao, P.D.; Zhao, P.P.; Zhao, W.R.; Zhao, Z.G.; Zhao, Z.Q.; Zheng, J.P.; Zheng, L.S.; Zheng, M.; Zheng, W.S.; Zheng, Z.P.; Zhong, G.P.; Zhou, G.P.; Zhou, H.S.; Zhou, J.; Zhou Li; Zhou Lin; Zhou, M.; Zhou, Y.S.; Zhou, Y.H.; Zhu, G.S.; Zhu, Q.M.; Zhu, S.G.; Zhu, Y.C.; Zhu, Y.S.; Zhuang, B.A.

    1994-01-01

    The Beijing Spectrometer (BES) is a general purpose solenoidal detector at the Beijing Electron Positron Collider (BEPC). It is designed to study exclusive final states in e + e - annihilations at the center of mass energy from 3.0 to 5.6 GeV. This requires large solid angle coverage combined with good charged particle momentum resolution, good particle identification and high photon detection efficiency at low energies. In this paper we describe the construction and the performance of BES detector. (orig.)

  1. Superlattice electroabsorption radiation detector

    International Nuclear Information System (INIS)

    Cooke, B.J.

    1993-06-01

    This paper provides a preliminary investigation of a new class of superlattice electroabsorption radiation detectors that employ direct optical modulation for high-speed, two-dimensional (2-D), high-resolution imaging. Applications for the detector include nuclear radiation measurements, tactical guidance and detection (laser radar), inertial fusion plasma studies, and satellite-based sensors. Initial calculations discussed in this paper indicate that a 1.5-μm (GaAlAs) multi-quantum-well (MQW) Fabry-Perot detector can respond directly to radiation of energies 1 eV to 10 KeV, and indirectly (with scattering targets) up through gamma, with 2-D sample rates on the order of 20 ps

  2. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  3. Highly heat removing radiation shielding material

    International Nuclear Information System (INIS)

    Asano, Norio; Hozumi, Masahiro.

    1990-01-01

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

  4. Shielding design of ITER pressure suppression system

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu

    2006-01-01

    The duct shield from streaming D-T neutrons has been designed for the ITER pressure suppression system. Streaming calculations are performed with the DUCT-III code for the region from the inlet of the pressure relief line to the rupture disk. Next, the neutron permeation through the shield is studied by Monte Carlo calculations with the MCNP code. It is found that 0.15 m thick iron shield is enough to suppress the permeating component from the outside. In addition, it is suggested that the volume of the shield can be reduced by about 30% if the optimized iron shield structure having localized thickness across intense permeation paths is employed to shield the pressure suppression line. (T.I.)

  5. A Scintillator Purification System for the Borexino Solar Neutrino Detector

    OpenAIRE

    Benziger, J.; Cadonati, L.; Calaprice, F.; Chen, M.; Corsi, A.; Dalnoki-Veress, F.; Fernholz, R.; Ford, R.; Galbiati, C.; Goretti, A.; Harding, E.; Ianni, Aldo; Ianni, Andrea; Kidner, S.; Leung, M.

    2007-01-01

    Purification of the 278 tons of liquid scintillator and 889 tons of buffer shielding for the Borexino solar neutrino detector was performed with a system that combined distillation, water extraction, gas stripping and filtration. The purification of the scintillator achieved unprecedented low backgrounds for the large scale liquid scintillation detector. This paper describes the principles of operation, design, construction and commissioning of the purification system, and reviews the require...

  6. Comparative Study of Radiation Shielding Parameters for Bismuth Borate Glasses

    OpenAIRE

    Kaundal, Rajinder Singh

    2016-01-01

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi2O3-(1-x) B2O3 where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of...

  7. Irrigoscopy - irrigography method, dosimetry and radiation shielding

    International Nuclear Information System (INIS)

    Zubanov, Z.; Kolarevic, G.

    1999-01-01

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

  8. Cage for shield-type support. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Harryers, W; Blumenthal, G; Irresberger, H

    1981-08-13

    A cage for shield-type support containing a fracture shield supported by a hydraulic stamp and a projecting roof bar was constructed in such a way that no cellular shirt is needed to timber the caved room. The roof bar which is linked at a joint axis at the face-side end of the fracture shield is formed at the face side as a multiply foldable bar. (HGOE).

  9. Influence of sampling properties of fast-waveform digitizers on neutron−gamma-ray, pulse-shape discrimination for organic scintillation detectors

    International Nuclear Information System (INIS)

    Flaska, Marek; Faisal, Muhammad; Wentzloff, David D.; Pozzi, Sara A.

    2013-01-01

    One of the most important questions to be answered with regard to digital pulse-shape discrimination (PSD) systems based on organic scintillators is: What sampling properties are required for a fast-waveform digitizer used for digitizing neutron/gamma-ray pulses, while an accurate PSD is desired? Answering this question is the main objective of this paper. Specifically, the paper describes the influence of the resolution and sampling frequency of a waveform digitizer on the PSD performance of organic scintillators. The results presented in this paper are meant to help the reader choosing a waveform digitizer with appropriate bit resolution and sampling frequency. The results presented here show that a 12-bit, 250-MHz digitizer is a good choice for applications that require good PSD performance. However, when more accurate PSD performance is the main requirement, this paper presents PSD figures of merit to qualify the impact of further increasing either sampling frequency or resolution of the digitizer

  10. Magnesium borate radiothermoluminescent detectors

    International Nuclear Information System (INIS)

    Kazanskaya, V.A.; Kuzmin, V.V.; Minaeva, E.E.; Sokolov, A.D.

    1974-01-01

    In the report the technology of obtaining polycrystalline magnesium borate activated by dysprosium is described briefly and the method of preparing the tabletted detectors from it is presented. The dependence of the light sum of the samples on the proportion of the components and on the sintering regime has shown that the most sensitive material is obtained at the proportion of boric anhydride and magnesium oxide 2.2-2.4 and at the dysprosium concentration about 1 milligram-atom per gram molecule of the base. The glow curve of such a material has a simple form with one peak the maximum of which is located at 190-200 0 C. The measurement of the main dosimetric characteristics of the magnesium borate tabletted detectors and the comparison with similar parmaeters of the lithium fluoride tabletted detectors have shown that at practically identical effective number the former detectors have the following substantial advantages: the sensitivity is ten-twenty times as large, they are substantially more technological on synthesis of the radiothermoluminophor and during the production of the tabletted detectors, they have a simple glow curve, they do not require the utilization of the thermocycling during the use. (author)

  11. Problems of the power plant shield optimization

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  12. Thermal design of top shield for PFBR

    International Nuclear Information System (INIS)

    Gajapathy, R.; Jalaludeen, S.; Selvaraj, A.; Bhoje, S.B.

    1988-01-01

    India's Liquid Metal Cooled Fast Breeder Reactor programme started with the construction of loop type 13MW(e) Fast Breeder Test Reactor (FBTR) which attained criticality in October 1985. With the experience of FBTR, the design work on pool type 500 MW(e) Prototype Fast Breeder Reactor (PFBR) which will be a forerunner for future commercial fast breeder reactors, has been started. The Top Shield forms the cover for the main vessel which contains the primary circuit. Argon cover gas separates the Top Shield from the free level of hot sodium pool (803K). The Top Shield which is of box type construction consists of control plug, two rotatable plugs and roof slab, assembled together, which provide biological shielding, thermal shielding and leak tight containment at the top of the main vessel. Heat is transferred from the sodium pool to the Top Shield through argon cover gas and through components supported by it and dipped in the sodium pool. The Top Shield should be maintained at the desired operating temperature by incorporating a cooling system inside it. Insulation may be provided below the bottom plate to reduce the heat load to the cooling system, if required. The thermal design of Top Shield consists of estimation of heat transfer to the Top Shield, selection of operating temperature, assessment of insulation requirement, design of cooling system and evaluation of transient temperature changes

  13. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente

    2006-01-01

    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 252 Cf isotopic neutron source. During calculations a detailed model for the 252 Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252 Cf 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)

  14. Hot Cell Window Shielding Analysis Using MCNP

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. Radiation shielding application of lead glass

    International Nuclear Information System (INIS)

    Nathuram, R.

    2017-01-01

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

  16. Radiation dose reduction by water shield

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  17. Testing of the KRI-developed Silicon PIN Radioxenon Detector

    International Nuclear Information System (INIS)

    Foxe, Michael P.; McIntyre, Justin I.

    2015-01-01

    Radioxenon detectors are used for the verification of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) in a network of detectors throughout the world called the International Monitoring System (IMS). The Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) Provisional Technical Secretariat (PTS) has tasked Pacific Northwest National Laboratory (PNNL) with testing a V.G. Khlopin Radium Institute (KRI) and Lares Ltd-developed Silicon PIN detector for radioxenon detection. PNNL measured radioxenon with the silicon PIN detector and determined its potential compared to current plastic scintillator beta cells. While the PNNL tested Si detector experienced noise issues, a second detector was tested in Russia at Lares Ltd, which did not exhibit the noise issues. Without the noise issues, the Si detector produces much better energy resolution and isomer peak separation than a conventional plastic scintillator cell used in the SAUNA systems in the IMS. Under the assumption of 1 cm 3 of Xe in laboratory-like conditions, 24-hr count time (12-hr count time for the SAUNA), with the respective shielding the minimum detectable concentrations for the Si detector tested by Lares Ltd (and a conventional SAUNA system) were calculated to be: 131m Xe - 0.12 mBq/m 3 (0.12 mBq/m 3 ); 133 Xe - 0.18 mBq/m 3 (0.21 mBq/m 3 ); 133m Xe - 0.07 mBq/m 3 (0.15 mBq/m 3 ); 135 Xe - 0.45 mBq/m 3 (0.67 mBq/m 3 ). Detection limits, which are one of the important factors in choosing the best detection technique for radioxenon in field conditions, are significantly better than for SAUNA-like detection systems for 131m Xe and 133m Xe, but similar for 133 Xe and 135 Xe. Another important factor is the amount of ''memory effect'' or carry over signal from one radioxenon measurement to the subsequent sample. The memory effect is reduced by a factor of 10 in the Si PIN detector compared to the current plastic scintillator cells. There is potential for further reduction with the

  18. Correction of rhodium detector signals for comparison to design calculations

    International Nuclear Information System (INIS)

    Judd, J.L.; Chang, R.Y.; Gabel, C.W.

    1989-01-01

    Rhodium detectors are used in many commercial pressurized water reactors PWRs [pressurized water reactor] as in-core neutron detectors. The signals from the detectors are the result of neutron absorption in 103 Rh and the subsequent beta decay of 104 Rh to 104 Pd. The rhodium depletes ∼1% per full-power month, so corrections are necessary to the detector signal to account for the effects of the rhodium depletion. These corrections result from the change in detector self-shielding with rhodium burnup and the change in rhodium concentration itself. Correction for the change in rhodium concentration is done by multiplication of the factor N(t)/N 0 , where N(t) is the rhodium concentration at time t and N 0 is the initial rhodium concentration. The calculation of the self-shielding factor is more complicated and is presented. A self-shielding factor based on the fraction of rhodium remaining was calculated with the CASMO-3 code. The results obtained from our comparisons of predicted and measured in-core detector signals show that the CASMO-3/SIMULATE-3 code package is an effective tool for estimating pin peaking and power distributions

  19. Vertex detectors

    International Nuclear Information System (INIS)

    Lueth, V.

    1992-07-01

    The purpose of a vertex detector is to measure position and angles of charged particle tracks to sufficient precision so as to be able to separate tracks originating from decay vertices from those produced at the interaction vertex. Such measurements are interesting because they permit the detection of weakly decaying particles with lifetimes down to 10 -13 s, among them the τ lepton and charm and beauty hadrons. These two lectures are intended to introduce the reader to the different techniques for the detection of secondary vertices that have been developed over the past decades. The first lecture includes a brief introduction to the methods used to detect secondary vertices and to estimate particle lifetimes. It describes the traditional technologies, based on photographic recording in emulsions and on film of bubble chambers, and introduces fast electronic registration of signals derived from scintillating fibers, drift chambers and gaseous micro-strip chambers. The second lecture is devoted to solid state detectors. It begins with a brief introduction into semiconductor devices, and then describes the application of large arrays of strip and pixel diodes for charged particle tracking. These lectures can only serve as an introduction the topic of vertex detectors. Time and space do not allow for an in-depth coverage of many of the interesting aspects of vertex detector design and operation

  20. Smoke detectors

    International Nuclear Information System (INIS)

    Macdonald, E.

    1976-01-01

    A smoke detector is described consisting of a ventilated ionisation chamber having a number of electrodes and containing a radioactive source in the form of a foil supported on the surface of the electrodes. This electrode consists of a plastic material treated with graphite to render it electrically conductive. (U.K.)

  1. Semiconductor Detectors

    International Nuclear Information System (INIS)

    Cortina, E.

    2007-01-01

    Particle detectors based on semiconductor materials are among the few devices used for particle detection that are available to the public at large. In fact we are surrounded by them in our daily lives: they are used in photoelectric cells for opening doors, in digital photographic and video camera, and in bar code readers at supermarket cash registers. (Author)

  2. Capillary detectors

    International Nuclear Information System (INIS)

    Konijn, J.; Winter, K.; Vilain, P.; Wilquet, G.; Fabre, J.P.; Kozarenko, E.; Kreslo, I.; Goldberg, J.; Hoepfner, K.; Bay, A.; Currat, C.; Koppenburg, P.; Frekers, D.; Wolff, T.; Buontempo, S.; Ereditato, A.; Frenkel, A.; Liberti, B.; Martellotti, G.; Penso, G.; Ekimov, A.; Golovkin, S.; Govorun, V.; Medvedkov, A.; Vasil'chenko, V.

    1998-01-01

    The option for a microvertex detector using glass capillary arrays filled with liquid scintillator is presented. The status of capillary layers development and possible read-out techniques for high rate environment are reported. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  5. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.

    1994-01-01

    Lead used for shielding is often surface contaminated with radioisotopes and is therefore a 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 Laboratory 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 50 tons and likely to grow substantially of planned decommissioning operations. Thus lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for contaminated lead is removing the superficial layer of contamination with an abrasive medium under pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a scaled-off area. The slurry of abrasive and particles of lead falls through a floor and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling

  6. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

    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

  7. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

    Lead used for shielding is often surface contaminated with radioisotopes and is therefore a 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 Laboratory 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 50 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 trader pressure. For lead, a mixture of alumina with water and air at about 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 sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of contaminated lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling

  8. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    Rennich, M.J.; Nelson, B.E.

    1994-01-01

    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

  9. Isotope effects on nuclear shielding

    International Nuclear Information System (INIS)

    Hansen, P.E.

    1983-01-01

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

  10. Continuous electrodeionization through electrostatic shielding

    International Nuclear Information System (INIS)

    Dermentzis, Konstantinos

    2008-01-01

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

  11. Compton suppression tests on Ge and BGO prototype detectors for GAMMASPHERE

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, A M [Australian National Univ., Canberra, ACT (Australia); Khoo, T L; Bleich, M E; Carpenter, M P; Ahmad, I; Janssens, R V.F.; Moore, E F [Argonne National Lab., IL (United States); Bearden, I G [Purdue Univ., Lafayette, IN (United States); Beene, J R; Lee, I Y [Oak Ridge National Lab., TN (United States)

    1992-08-01

    In this paper, we report on measurements of the Compton suppression and overall P/T ratio of two Ge detectors in a BGO shield of the honeycomb pattern. These were the first prototype CSG detector assemblies for GAMMASPHERE. A more detailed description of these results will be published later. (author). 4 refs., 3 figs.

  12. Electron, electron-bremsstrahlung and proton depth-dose data for space-shielding applications

    Science.gov (United States)

    Seltzer, S. M.

    1979-01-01

    A data set has been developed, consisting of depth-dose distributions for omni-directional electron and proton fluxes incident on aluminum shields. The principal new feature of this work is the accurate treatment, based on detailed Monte Carlo calculations, of the electron-produced bremsstrahlung component. Results covering the energy region of interest in space-shielding calculations have been obtained for the absorbed dose (a) as a function of depth in a semi-infinite medium, (b) at the edge of slab shields, and (c) at the center of a solid sphere. The dose to a thin tissue-equivalent detector was obtained as well as that in aluminum. Various results and comparisons with other work are given.

  13. Comparison of experimental and calculated shielding factors for modular buildings in a radioactive fallout scenario

    DEFF Research Database (Denmark)

    Hinrichsen, Yvonne; Finck, Robert; Östlund, Karl

    2018-01-01

    building used was a standard prefabricated structure obtained from a commercial manufacturer. Four reference positions for the gamma radiation detectors were used inside the building. Theoretical dose rate calculations were performed using the Monte Carlo code MCNP6, and additional calculations were......Experimentally and theoretically determined shielding factors for a common light construction dwelling type were obtained and compared. Sources of the gamma-emitting radionuclides 60Co and 137Cs were positioned around and on top of a modular building to represent homogeneous fallout. The modular...... performed that compared the shielding factor for 137Cs and 134Cs. This work demonstrated the applicability of using MCNP6 for theoretical calculations of radioactive fallout scenarios. Furthermore, the work showed that the shielding effect for modular buildings is almost the same for 134Cs as for 137Cs....

  14. A Monte Carlo study for the shielding of γ backgrounds induced by radionuclides for CDEX

    International Nuclear Information System (INIS)

    Li Lei; Tang Changjian; Yue Qian; Cheng Jianping; Kang Kejun; Li Jianmin; Li Jin; Li Yulan; Li Yuanjing; Ma Hao; Xue Tao; Zeng Zhi; Wong, H.T.

    2011-01-01

    The CDEX (China Dark matter EXperiment) Collaboration will carry out a direct search for WIMPs (Weakly Interacting Massive Particles) using an Ultra-Low Energy Threshold High Purity Germanium (ULE-HPGe) detector at the CJPL (China JinPing deep underground Laboratory). A complex shielding system was designed to reduce backgrounds and a detailed GEANT4 Monte Carlo simulation was performed to study the achievable reduction of γ rays induced by radionuclides and neutron backgrounds by D(γ,n)p reaction. Furthermore, the upper level of allowed radio purity of shielding materials was estimated under the constraint of the expected goal. Compared with the radio purity reported by other low-background rare-event experiments, it indicates that the shielding used in the CDEX can be made out of materials with obtainable radiopurity. (authors)

  15. Development and installation of an automatic sample changer for neutron activation analysis

    International Nuclear Information System (INIS)

    Domienikan, Claudio; Lapolli, Andre L.; Schoueri, Roberto M.; Moreira, Edson G.; Vasconcellos, Marina B.A.

    2013-01-01

    A Programmable and Automatic Sample Changer was built and installed at the Neutron Activation Analysis Laboratory of the Nuclear and Energy Research Institute - IPEN-CNEN/SP, Brazil. This Automatic Sample Changer allows the fully automated measurement of up to 25 samples in one run. Basically it consists of an electronic circuit and C++ program that controls the positioning of a sample holder in two axes of motion (X and Y). Each sample is transported and positioned, one by one, inside the shielding coupled to a high-purity germanium (HPGe) radiation detector. A Canberra DSA-1000 Multichannel Analyzer coupled to the Genie 2000 software performs the data acquisition for analysis of the samples. When the counting is finished the results are saved in a hard disk of a PC computer. The sample is brought back by the sample holder to its initial position, and the next sample is carried to the shielding. The Sample Changer was designed and constructed at IPEN-CNEN/SP by employing national components and expertise. (author)

  16. Unresolved resonance self shielding calculation: causes and importance of discrepancies

    International Nuclear Information System (INIS)

    Ribon, P.; Tellier, H.

    1986-09-01

    To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs

  17. Unresolved resonance self shielding calculation: causes and importance of discrepancies

    International Nuclear Information System (INIS)

    Ribon, P.; Tellier, H.

    1986-01-01

    To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, the parameters of each level are not known; only the average parameters. Therefore the authors simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the x 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, the authors survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors

  18. Gamma radiation shielding analysis of lead-flyash concretes

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. A scintillator purification system for the Borexino solar neutrino detector

    Science.gov (United States)

    Benziger, J.; Cadonati, L.; Calaprice, F.; Chen, M.; Corsi, A.; Dalnoki-Veress, F.; Fernholz, R.; Ford, R.; Galbiati, C.; Goretti, A.; Harding, E.; Ianni, Aldo; Ianni, Andrea; Kidner, S.; Leung, M.; Loeser, F.; McCarty, K.; McKinsey, D.; Nelson, A.; Pocar, A.; Salvo, C.; Schimizzi, D.; Shutt, T.; Sonnenschein, A.

    2008-03-01

    Purification of the 278 tons of liquid scintillator and 889 tons of buffer shielding for the Borexino solar neutrino detector is performed with a system that combines distillation, water extraction, gas stripping, and filtration. This paper describes the principles of operation, design, and construction of that purification system, and reviews the requirements and methods to achieve system cleanliness and leak-tightness.

  20. Comparison among Models to Estimate the Shielding Effectiveness Applied to Conductive Textiles

    Directory of Open Access Journals (Sweden)

    Alberto Lopez

    2013-01-01

    Full Text Available The purpose of this paper is to present a comparison among two models and its measurement to calculate the shielding effectiveness of electromagnetic barriers, applying it to conductive textiles. Each one, models a conductive textile as either a (1 wire mesh screen or (2 compact material. Therefore, the objective is to perform an analysis of the models in order to determine which one is a better approximation for electromagnetic shielding fabrics. In order to provide results for the comparison, the shielding effectiveness of the sample has been measured by means of the standard ASTM D4935-99.