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)

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.

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

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

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)

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)

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

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.

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.

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)

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.

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

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)

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.

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.

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

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

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.

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.

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.

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.

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)

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

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

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.

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.

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.

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)

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)

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

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)

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.

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

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)

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.

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)

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)

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)

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

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)

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

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)

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)

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

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.

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...

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)

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

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.

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)

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).

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.)

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

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.

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)

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.

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

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)

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.

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.

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.

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

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)

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

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.

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.)

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.

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.

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

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.)

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...

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)

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.

Background intercomparison with escape-suppressed germanium detectors in underground mines

Energy Technology Data Exchange (ETDEWEB)

Szuecs, Tamas; Bemmerer, Daniel [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany)

2014-07-01

A key requirement for underground nuclear astrophysics experiments is the very low background level in germanium detectors underground. The reference for these purposes is the world's so far only underground accelerator laboratory for nuclear astrophysics, LUNA. LUNA is located deep underground in the Gran Sasso laboratory in Italy, shielded from cosmic rays by 1400 m of rock. The background at LUNA was studied in detail using an escape-suppressed Clover-type HPGe detector. Exactly the same detector was subsequently transported to the Felsenkeller underground laboratory in Dresden, shielded by 45 m of rock, and the background was shown to be only a factor of three higher than at LUNA when comparing the escape-suppressed spectra, with interesting consequences for underground nuclear astrophysics. As the next step of a systematic study of the effects of a combination of active and passive shielding on the cosmic ray induced background, this detector is now being brought to the ''Reiche Zeche'' mine in Freiberg/Sachsen, shielded by 150 m of rock. The data from the Freiberg measurement are shown and discussed.

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.)

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

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

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....

Monte Carlo simulation of determining porosity by using dual gamma detectors

International Nuclear Information System (INIS)

Zhang Feng; Liu Juntao; Yu Huawei; Yuan Chao; Jia Yan

2013-01-01

Current formation elements spectroscopy logging technology utilize 241 Am-Be neutron source and single BGO detector to determine elements contents. It plays an important role in mineral analysis and lithology identification of unconventional oil and gas exploration, but information measured is relatively ld. Measured system based on 241 Am-Be neutron and dual detectors can be developed to realize the measurement of elements content as well as determine neutron gamma porosity by using ratio of gamma count between near and far detectors. Calculation model is built by Monte Carlo method to study neutron gamma porosity logging response with different spacing and shields. And it is concluded that measuring neutron gamma have high counts and good statistical property contrasted with measuring thermal neutron, but the sensitivity of porosity decrease. Sensitivity of porosity will increase as the spacing of dual detector increases. Spacing of far and near detectors should be around 62 cm and 35 cm respectively. Gamma counts decrease and neutron gamma porosity sensitivity increase when shield is fixed between neutron and detector. The length of main shield should be greater than 10 cm and associated shielding is about 5 cm. By Monte Carlo Simulation study, the result provides technical support for determining porosity in formation elements spectroscopy logging using 241 Am-Be neutron and gamma detectors. (authors)

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

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.

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.

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.

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

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...

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...

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.

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.

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.

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

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

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

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.

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.

Measuring space radiation shielding effectiveness

Directory of Open Access Journals (Sweden)

Bahadori Amir

2017-01-01

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

Measuring space radiation shielding effectiveness

Science.gov (United States)

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

2017-09-01

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

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

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

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

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)

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.)

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.

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)

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

Measurements and simulation-based optimization of TIGRESS HPGe detector array performance

International Nuclear Information System (INIS)

Schumaker, M.A.

2005-01-01

TIGRESS is a new γ-ray detector array being developed for installation at the new ISAC-II facility at TRIUMF in Vancouver. When complete, it will consist of twelve large-volume segmented HPGe clover detectors, fitted with segmented Compton suppression shields. The combined operation of prototypes of both a TIGRESS detector and a suppression shield has been tested. Peak-to-total ratios, relative photopeak efficiencies, and energy resolution functions have been determined in order to characterize the performance of TIGRESS. This information was then used to refine a GEANT4 simulation of the full detector array. Using this simulation, methods to overcome the degradation of the photopeak efficiency and peak-to-total response that occurs with high γ-ray multiplicity events were explored. These methods take advantage of the high segmentation of both the HPGe clovers and the suppression shields to suppress or sum detector interactions selectively. For a range of γ-ray energies and multiplicities, optimal analysis methods have been determined, which has resulted in significant gains in the expected performance of TIGRESS. (author)

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

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.

Progress toward a Km-scale neutrino detector in the deep ocean

Energy Technology Data Exchange (ETDEWEB)

Stokstad, R.G.

1997-11-01

The best particles for observing distant objects are photons and neutrinos. Because of the neutrino`s weak interaction cross section, detectors suitable for astronomy must be very large and well shielded from cosmic rays. Eventually, a detector with the order of a square km of effective area will be needed for systematic observations of distant point sources such as active galactic nuclei. Prototype detectors are currently being developed at several sites in the ocean, at Lake Baikal, and in Antarctica. This talk summarizes the status of the projects that use the deep ocean for the detector medium and shielding: DUMAND, NESTOR and ANTARES. Technical developments will be needed for a future km-scale detector; progress on one of these, a digital electronic system, is also described.

Progress toward a Km-scale neutrino detector in the deep ocean

International Nuclear Information System (INIS)

Stokstad, R.G.

1997-11-01

The best particles for observing distant objects are photons and neutrinos. Because of the neutrino's weak interaction cross section, detectors suitable for astronomy must be very large and well shielded from cosmic rays. Eventually, a detector with the order of a square km of effective area will be needed for systematic observations of distant point sources such as active galactic nuclei. Prototype detectors are currently being developed at several sites in the ocean, at Lake Baikal, and in Antarctica. This talk summarizes the status of the projects that use the deep ocean for the detector medium and shielding: DUMAND, NESTOR and ANTARES. Technical developments will be needed for a future km-scale detector; progress on one of these, a digital electronic system, is also described

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

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

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

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)

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)

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

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

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)

An ultra-low-background detector for axion searches

International Nuclear Information System (INIS)

Aune, S; Ferrer Ribas, E; Giomataris, I; Mols, J P; Papaevangelou, T; Dafni, T; Lacarra, J Galan; Iguaz, F J; Irastorza, I G; Morales, J; Ruz, J; Tomas, A; Fanourakis, G; Geralis, T; Kousouris, K; Vafeiadis, T

2009-01-01

A low background Micromegas detector has been operating in the CAST experiment at CERN for the search of solar axions since the start of data taking in 2002. The detector, made out of low radioactivity materials, operated efficiently and achieved a very low level of background (5x10 -5 keV -1 -cm -2 -s -1 ) without any shielding. New manufacturing techniques (Bulk/Microbulk) have led to further improvement of the characteristics of the detector such as uniformity, stability and energy resolution. These characteristics, the implementation of passive shielding and the improvement of the analysis algorithms have dramatically reduced the background level (2x10 -7 keV -1 -cm -2 |s -1 ), improving thus the overall sensitivity of the experiment and opening new possibilities for future searches.

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.)

Photon spectrum behind biological shielding of the LVR-15 research reactor

Energy Technology Data Exchange (ETDEWEB)

Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

2011-07-01

The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

Wood efficiency as passive shield for particles and photons of 0,5 to 4,6 MeV between 940 and 377 mb

International Nuclear Information System (INIS)

Aguiar, O.D. de; Nordemann, D.J.R.

1986-01-01

A pair of scintillators NaI (Tl), 4'x4', one completly shielded with 20 g/cm 2 of high density wood (1.3 g/cm 3 ) and the other unshielded, has been flown aboard a Bandeirante aircraft over the region of Sao Jose dos Campos (23 0 14'S, 314 0 9'E), up to altitudes of 25000 feet (377mb). The spectra of the detectors with and without the wood shield are similar, and the spectral indices indicate that they are function of atmospheric depth in the range 0.6-2.4 MeV. Between 2.4 and 4.6 MeV the shielded detector presented a counting rate which is 20% lower than the counting rate obtained by the unshielded detector. In this same energy range the counting rate of the shielded detector is equal to the counting rate of the unshielded detector located at an atmospherical depth of 35 g/cm 2 higher. The Attenuation length showed a decrease with the energy in the 0.65 to 4.6 MeV range and an increase above this energy. This increase is, probably, due to the strong influence of muons and electrons in this range. (Author) [pt

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

SU-F-T-490: Separating Effects Influencing Detector Response in Small MV Photon Fields

Energy Technology Data Exchange (ETDEWEB)

Wegener, S; Sauer, O [University of Wuerzburg, Wuerzburg (Germany)

2016-06-15

Purpose: Different detector properties influence their responses especially in field sizes below the lateral electron range. Due to the finite active volume, the detector density and electron perturbation at other structural parts, the response factor is in general field size dependent. We aimed to visualize and separate the main effects contributing to detector behavior for a variety of detector types. This was achieved in an experimental setup, shielding the field center. Thus, effects caused by scattered radiation could be examined separately. Methods: Signal ratios for field sizes down to 8 mm (SSD 90 cm, water depth 10 cm) of a 6MV beam from a Siemens Primus LINAC were recorded with several detectors: PTW microDiamond and PinPoint ionization chamber, shielded diodes (PTW P-60008, IBA PFD and SNC Edge) and unshielded diodes (PTW E-60012 and IBA SFD). Measurements were carried out in open fields and with an aluminum pole of 4 mm diameter as a central block. The geometric volume effect was calculated from profiles obtained with Gafchromic EBT3 film, evaluated using FilmQA Pro software (Ashland, USA). Results: Volume corrections were 1.7% at maximum. After correction, in small open fields, unshielded diodes showed a lower response than the diamond, i.e. diamond detector over-response seems to be higher than that for unshielded diodes. Beneath the block, this behavior was amplified by a factor of 2. For the shielded diodes, the overresponse for small open fields could be confirmed. However their lateral response behavior was strongly type dependent, e.g. the signal ratio dropped from 1.02 to 0.98 for the P-60008 diode. Conclusion: The lateral detector response was experimentally examined. Detector volume and density alone do not fully account for the field size dependence of detector response. Detector construction details play a major role, especially for shielded diodes.

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

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

A balloon-borne solid state cosmic X-ray detector

International Nuclear Information System (INIS)

Proctor, R.; Pietsch, W.; Reppin, C.

1982-01-01

On 9th May 1980 a MPI/AIT hard X-ray balloon payload successfully observed numerous cosmic X-ray sources. The payload consisted of a 2400 cm 2 Phoswich detector and a 114 cm 2 solid state detector. The solid state detector is described in this report. It consists of six intrinsic germanium planar crystals in a vacuum cryostat cooled by liquid nitrogen. The detector operates in the hard X-ray energy range of 20-150 keV and had in-flight a mean energy resolution of 2.75 keV at 60 keV. A hexagonal molybdenum collimator defined the field of view as approximately 4 0 fwhm. A CsI(Na) and plastic active shield and passive shielding provided background rejection. Mean background values of 1.3 X 10 -3 counts/(sec x cm 2 x keV) at 60 keV were obtained. (orig.)

TH-CD-201-02: A Monte Carlo Investigation of a Novel Detector Arrangement for the Energy Spectrum Measurement of a 6MV Linear Accelerator

Energy Technology Data Exchange (ETDEWEB)

Taneja, S; Bartol, L; Culberson, W; DeWerd, L [School of Medicine and Public Health, University of Wisconsin-Madison, Madison, WI (United States)

2016-06-15

Purpose: Direct measurement of the energy spectrum of a 6MV linear accelerator has not been successful due to the high fluence rate, high energy nature of these photon beams. Previous work used a Compton Scattering (CS) spectrometry setup with a shielded spectrometer for spectrum measurements. Despite substantial lead shielding, excessive pulse pile-up was seen. MCNP6 transport code was used to investigate the feasibility and effectiveness of performing measurements using a novel detector setup. Methods: Simulations were performed with a shielded high-purity germanium (HPGe) semiconductor detector placed in the accelerator vault’s maze, with a 2 cm diameter collimator through a 92 cm thick concrete wall. The detector was positioned 660 cm from a scattering rod (placed at isocenter) at an angle of 45° relative to the central axis. This setup was compared with the shielded detector positioned in the room, 200 cm from the scattering rod at the same CS angle. Simulations were used to determine fluence contributions from three sources: (1) CS photons traveling through the collimator aperture, the intended signal, (2) CS scatter photons penetrating the detector shield, and (3) room-scattered photons penetrating the detector shield. Variance reduction techniques including weight windows, DXTRAN spheres, forced collisions, and energy cutoffs were used. Results: Simulations showed that the number of pulses per starting particle from an F8 detector tally for the intended signal decreased by a factor of 10{sup 2} when moving the detector out of the vault. This reduction in signal was amplified for the unwanted scatter signal which decreased by up to a factor of 10{sup 9}. Conclusion: This work used MCNP6 to show that using a vault wall to shield unwanted scatter and increasing isocenter-to-detector distance reduces unwanted fluence to the detector. This study aimed to provide motivation for future experimental work using the proposed setup.

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.

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)

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)

An ultra-low-background detector for axion searches

Energy Technology Data Exchange (ETDEWEB)

Aune, S; Ferrer Ribas, E; Giomataris, I; Mols, J P; Papaevangelou, T [IRFU, Centre d' Etudes de Saclay, Gif sur Yvette CEDEX (France); Dafni, T; Lacarra, J Galan; Iguaz, F J; Irastorza, I G; Morales, J; Ruz, J; Tomas, A [Instituto de Fisica Nuclear y Altas EnergIas, Zaragoza (Spain); Fanourakis, G; Geralis, T; Kousouris, K [Institute of Nuclear Physics, NCSR Demokritos, Athens (Greece); Vafeiadis, T, E-mail: Thomas.Papaevangelou@cern.c [Physics Department, Aristotle University, Thessaloniki (Greece)

2009-07-01

A low background Micromegas detector has been operating in the CAST experiment at CERN for the search of solar axions since the start of data taking in 2002. The detector, made out of low radioactivity materials, operated efficiently and achieved a very low level of background (5x10{sup -5} keV{sup -1}-cm{sup -2}-s{sup -1}) without any shielding. New manufacturing techniques (Bulk/Microbulk) have led to further improvement of the characteristics of the detector such as uniformity, stability and energy resolution. These characteristics, the implementation of passive shielding and the improvement of the analysis algorithms have dramatically reduced the background level (2x10{sup -7} keV{sup -1}-cm{sup -2}|s{sup -1}), improving thus the overall sensitivity of the experiment and opening new possibilities for future searches.

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

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

Application of a simple analytical model to estimate effectiveness of radiation shielding for neutrons

International Nuclear Information System (INIS)

Frankle, S.C.; Fitzgerald, D.H.; Hutson, R.L.; Macek, R.J.; Wilkinson, C.A.

1993-01-01

Neutron dose equivalent rates have been measured for 800-MeV proton beam spills at the Los Alamos Meson Physics Facility. Neutron detectors were used to measure the neutron dose levels at a number of locations for each beam-spill test, and neutron energy spectra were measured for several beam-spill tests. Estimates of expected levels for various detector locations were made using a simple analytical model developed for 800-MeV proton beam spills. A comparison of measurements and model estimates indicates that the model is reasonably accurate in estimating the neutron dose equivalent rate for simple shielding geometries. The model fails for more complicated shielding geometries, where indirect contributions to the dose equivalent rate can dominate

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)

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

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.)

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.)

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

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 %

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)

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.)

Study of the use of a Phoswich detector for aerial radiometric surveys

International Nuclear Information System (INIS)

Schneid, E.J.; Lagin, L.J.

1981-09-01

Large volume, actively shielded Phoswich detector configurations were investigated for application in airborne radiometric survey systems in order to increase the sensitivity to ground-level potassium, uranium, and thorium compared to conventional NaI detectors of similar sizes. One Phoswich configuration was fabricated and flight tested. The flight measurements with the Phoswich detector and a conventional NaI detector utilized the Bendix Field Engineering Corporation - US Department of Energy/Grand Junction Office Walker Field Aerial Radiometric Survey Calibration Facility, the DOE dynamic test range, and a region in the southwest portion of the Lubbock, Texas, Quadrangle. The tests demonstrated that the Phoswich detector can function in an operational environment and does provide greater sensitivity to ground-level potassium, uranium, and thorium concentration than the conventional NaI detector. The multicrystal configuration for the Phoswich detector provided the capability to simultaneously measure the atmospheric radon background without the need of heavy lead shields. The flight tests indicated that the Phoswich detector has directional discrimination capability not available with conventional NaI detector systems. This directional capability can be used to distinguish the location, relative to the aircraft, of localized strong sources of radiation

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)

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...

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)

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.

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.)

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.

Measured neutron beam line shielding effectiveness of several iron/polyethylene configurations

International Nuclear Information System (INIS)

Legate, G.L.; Howe, M.L.; Mundis, R.L.

1988-01-01

Neutron and gamma-ray leakage measurements were taken at various stages of shield construction of neutron flight path 5 (the Lash-up flight path) at LANSCE, to compare the relative effectiveness of several configurations. Dose equivalent rates were determined for three categories: ''low-energy neutrons'', below 20 MeV; ''high- energy neutrons'', above 20 MeV; and gamma rays, as measured by hand-held survey instruments. The low energy neutrons were measured by activation of an indium foil in a paraffin-filled cadmium canister, sized to be generally insensitive above 20 MeV. High-energy neutrons were measured by (n,2n) production of Carbon 11 in a plastic scintillator with a 20-MeV threshold. Thermal neutrons were not measured at the shield-leakage test points. Room-scattered neutrons were observed by Albatross IV detector readings, which were taken beside the shield as a measure of variation of room background as the shield configuration changed. 1 fig., 1 tab

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)

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.

Astroparticle physics with a customized low-background broad energy Germanium detector

Energy Technology Data Exchange (ETDEWEB)

Aalseth, Craig E.; Amman, M.; Avignone, Frank T.; Back, Henning O.; Barabash, Alexander S.; Barbeau, P. S.; Bergevin, M.; Bertrand, F.; Boswell, M.; Brudanin, V.; Bugg, William; Burritt, Tom H.; Busch, Matthew; Capps, Greg L.; Chan, Yuen-Dat; Collar, J. I.; Cooper, R. J.; Creswick, R.; Detwiler, Jason A.; Diaz, J.; Doe, Peter J.; Efremenko, Yuri; Egorov, Viatcheslav; Ejiri, H.; Elliott, Steven R.; Ely, James H.; Esterline, James H.; Farach, H. A.; Fast, James E.; Fields, N.; Finnerty, P.; Fujikawa, Brian; Fuller, Erin S.; Gehman, Victor M.; Giovanetti, G. K.; Guiseppe, Vincente; Gusey, K.; Hallin, A. L.; Harper, Gregory; Hazama, R.; Henning, Reyco; Hime, Andrew; Hoppe, Eric W.; Hossbach, Todd W.; Howe, M. A.; Johnson, R. A.; Keeter, K.; Keillor, Martin E.; Keller, C.; Kephart, Jeremy D.; Kidd, Mary; Knecht, A.; Kochetov, Oleg; Konovalov, S.; Kouzes, Richard T.; Leviner, L.; Loach, J. C.; Luke, P.; MacMullin, S.; Marino, Michael G.; Martin, R. D.; Mei, Dong-Ming; Miley, Harry S.; Miller, M. L.; Mizouni, Leila; Myers, Allan W.; Nomachi, Masaharu; Orrell, John L.; Peterson, David; Phillips, D.; Poon, Alan; Prior, Gersende; Qian, J.; Radford, D. C.; Rielage, Keith; Robertson, R. G. H.; Rodriguez, Larry; Rykaczewski, Krzysztof P.; Salazar, Harold; Schubert, Alexis G.; Shima, T.; Shirchenko, M.; Steele, David; Strain, J.; Swift, Gary; Thomas, K.; Timkin, V.; Tornow, W.; Van Wechel, T. D.; Vanyushin, I.; Varner, R. L.; Vetter, Kai; Wilkerson, J. F.; Wolfe, B. A.; Xiang, W.; Yakushev, E.; Yaver, Harold; Young, A.; Yu, Chang-Hong; Yumatov, Vladimir; Zhang, C.; Zimmerman, S.

2011-10-01

The Majorana Collaboration is building the Majorana Demonstrator, a 60 kg array of high purity germanium detectors housed in an ultra-low background shield at the Sanford Underground Laboratory in Lead, SD. The Majorana Demonstrator will search for neutrinoless double-beta decay of 76Ge while demonstrating the feasibility of a tonne-scale experiment. It may also carry out a dark matter search in the 1-10 GeV/c² mass range. We have found that customized Broad Energy Germanium (BEGe) detectors produced by Canberra have several desirable features for a neutrinoless double-beta decay experiment, including low electronic noise, excellent pulse shape analysis capabilities, and simple fabrication. We have deployed a customized BEGe, the Majorana Low-Background BEGe at Kimballton (MALBEK), in a low-background cryostat and shield at the Kimballton Underground Research Facility in Virginia. This paper will focus on the detector characteristics and measurements that can be performed with such a radiation detector in a low-background environment.

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

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

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

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

A Muon Tomography Station with GEM Detectors for Nuclear Threat Detection

Science.gov (United States)

Staib, Michael; Gnanvo, Kondo; Grasso, Leonard; Hohlmann, Marcus; Locke, Judson; Costa, Filippo; Martoiu, Sorin; Muller, Hans

2011-10-01

Muon tomography for homeland security aims at detecting well-shielded nuclear contraband in cargo and imaging it in 3D. The technique exploits multiple scattering of atmospheric cosmic ray muons, which is stronger in dense, high-Z nuclear materials, e.g. enriched uranium, than in low-Z and medium-Z shielding materials. We have constructed and operated a compact Muon Tomography Station (MTS) that tracks muons with six to ten 30 cm x 30 cm Triple Gas Electron Multiplier (GEM) detectors placed on the sides of a 27-liter cubic imaging volume. The 2D strip readouts of the GEMs achieve a spatial resolution of ˜130 μm in both dimensions and the station is operated at a muon trigger rate of ˜20 Hz. The 1,536 strips per GEM detector are read out with the first medium-size implementation of the Scalable Readout System (SRS) developed specifically for Micro-Pattern Gas Detectors by the RD51 collaboration at CERN. We discuss the performance of this MTS prototype and present experimental results on tomographic imaging of high-Z objects with and without shielding.

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.)

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.)

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

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

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)

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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.

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)

Safety analyses in support of neutron detector calibration operations at JET

Energy Technology Data Exchange (ETDEWEB)

Stankunas, G., E-mail: gediminas@mail.lei.lt [EURATOM-LEI Association, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Syme, D.B.; Popovichev, S. [EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, S. [EURATOM-VR Association, Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Batistoni, P. [JET-EFDA Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-ENEA Association, Via E. Fermi, 40, 00044 Frascati (Italy)

2014-10-15

Highlights: •Neutron calculations to evaluate the dose rate leakage from the shields which contain the neutron source. •The differences on calculated dose rates using different flux-to-dose conversion factors have been investigated. •The experimental values were compared to the MCNPX calculations. -- Abstract: Neutron detectors in fusion devices need to be calibrated to provide the absolute neutron yield and the fusion power produced in fusion reactions. A new in situ calibration of the JET neutron detectors was recently performed using a {sup 252}Cf neutron source with intensity of about 2.7 × 10{sup 8} n/s. The source was delivered to the JET facility within a transport flask and the surface radiation levels must fall within transport regulations. Some contingency scenarios required transfer of the source into special shields: the operational shield and the auxiliary shield. In this paper we describe the neutron calculations that have been carried out to evaluate the dose rate leakage from the shields which may contain the neutron source. The calculations have been performed using accurate modelling of the neutron and gamma ray emission from the {sup 252}Cf source, and from the three shields. The differences on calculated dose rates deriving from the use of different flux-to-dose conversion factors have also been investigated. A comparison of dose rates calculated and measured is presented from the bare source (in cell) and with the source within its transport flask.

Contribution of 210Pb bremsstrahlung to the background of lead shielded gamma spectrometers

International Nuclear Information System (INIS)

Mrda, D.; Bikit, I.; Veskovic, M.; Forkapic, S.

2007-01-01

Lead, which is often used as a shielding material, contains 210 Pb (T 1/2 =22.3 y). The 46.54 keV γ-intensity of 210 Pb can be easily reduced by an inner lining, but the bremsstrahlung caused by the β-decay of its daughter, 210 Bi, with a maximal electron energy of 1.16 MeV, will contribute to the gamma detector background. The spectrum of this bremsstrahlung is calculated by numerically fitting the β-spectrum and integrating the Koch-Motz formula. The absorption of the bremsstrahlung in the lead and detection efficiencies for the HPGe detector are calculated by the effective solid angle algorithm, using corrections for the photopeak/Compton ratio of cross-sections in Ge. By comparison with the measured background spectrum, it is shown that, for the lead with 25 Bq/kg of 210 Pb up to 500 keV of gamma spectrum, the bremsstrahlung contribution to the background is about 20% for our surface-based detector system. Also, we compared our calculations with a Monte Carlo simulation of another detector system with a shield containing 1 Bq/kg of 210 Pb and found that our analytical method gives a value of roughly two times higher than the Monte Carlo one for the total bremsstrahlung contribution. The quality of the analytical semi-empirical method is proved by the reasonable agreement with the experimental results published

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.

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.)

Development and verification of signal processing system of avalanche photo diode for the active shields onboard ASTRO-H

Energy Technology Data Exchange (ETDEWEB)

Ohno, M., E-mail: ohno@hep01.hepl.hiroshima-u.ac.jp [Department of Physical Sciences, Hiroshima University, Hiroshima 739-8526 (Japan); Kawano, T.; Edahiro, I.; Shirakawa, H.; Ohashi, N.; Okada, C.; Habata, S.; Katsuta, J.; Tanaka, Y.; Takahashi, H.; Mizuno, T.; Fukazawa, Y. [Department of Physical Sciences, Hiroshima University, Hiroshima 739-8526 (Japan); Murakami, H.; Kobayashi, S.; Miyake, K.; Ono, K.; Kato, Y.; Furuta, Y.; Murota, Y.; Okuda, K. [Department of Physics, University of Tokyo, Tokyo 113-0033 (Japan); and others

2016-09-21

The hard X-ray Imager and Soft Gamma-ray Detector onboard ASTRO-H demonstrate high sensitivity to hard X-ray (5–80 keV) and soft gamma-rays (60–600 keV), respectively. To reduce the background, both instruments are actively shielded by large, thick Bismuth Germanate scintillators. We have developed the signal processing system of the avalanche photodiode in the BGO active shields and have demonstrated its effectiveness after assembly in the flight model of the HXI/SGD sensor and after integration into the satellite. The energy threshold achieved is about 150 keV and anti-coincidence efficiency for cosmic-ray events is almost 100%. Installed in the BGO active shield, the developed signal processing system successfully reduces the room background level of the main detector. - Highlights: • A detail of development of signal processing system for ASTRO-H is presented. • Digital filer with FPGA instead of discrete analog circuit is applied. • Expected performance is verified after integration of the satellite.

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

International Nuclear Information System (INIS)

Goebelbecker, Hans-Juergen

2008-01-01

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

Noise propagation issues in Belle II pixel detector power cable

Science.gov (United States)

Iglesias, M.; Arteche, F.; Echeverria, I.; Pradas, A.; Rivetta, C.; Moser, H.-G.; Kiesling, C.; Rummel, S.; Arcega, F. J.

2018-04-01

The vertex detector used in the upgrade of High-Energy physics experiment Belle II includes DEPFET pixel detector (PXD) technology. In this complex topology the power supply units and the front-end electronics are connected through a PXD power cable bundle which may propagate the output noise from the power supplies to the vertex area. This paper presents a study of the propagation of noise caused by power converters in the PXD cable bundle based on Multi-conductor Transmission Line (MTL) theory. The work exposes the effect of the complex cable topology and shield connections on the noise propagation, which has an impact on the requirements of the power supplies. This analysis is part of the electromagnetic compatibility based design focused on functional safety to define the shield connections and power supply specifications required to ensure the successful integration of the detector and, specifically, to achieve the designed performance of the front-end electronics.

Developments in gamma-ray spectrometry: systems, software, and methods-II. 3. Low-Energy Gamma-Ray Spectrometry Using a Compton-Suppressed Telescope Detector

International Nuclear Information System (INIS)

Sigg, R.A.; DiPrete, D.P.

2001-01-01

the new detector system has an ambient background count rate in the 30- to 600-keV region that is only 1/20 that of an earlier large area, low-energy photon spectrometer. For low-activity samples, whose counting backgrounds are dominated by natural background, minimum detectable activities are improved by a factor of ∼4. This improvement is attributable to a combination of factors including fabrication from low-background materials, cryostat geometry, passive and active shielding, and a lower ambient background location. Measurements using the newer system with and without gating reveal that most of this background reduction can be credited to active vetoing provided by guard detectors. Further improvements are anticipated as a nitrogen purge capability (to exclude radon from the shield) is implemented. An earlier paper (Ref. 2) described radiochemical separations and counting for 129 I, a long-lived fission product of interest at SRS because of the site's processing of spent nuclear fuels. While such radiochemical separations are necessary for many low-level analyses of 129 I, some low-density samples contain sufficient activity to allow 129 I analysis without chemistry if Compton-scattered interferences from other radionuclides are removed electronically. Tests on such a sample show that the sodium iodide and rear germanium anti-coincident guard detectors suppress the Compton continuum in the region near 129 I's 39-keV peak by a factor of ∼4. Interfering activities in the sample included 60 Co, 152 Eu, 154 Eu, and 241 Am. The Compton-suppression ratio is a function of gamma-ray energy; it improves as gamma ray energy increases until the best suppression (a factor of ∼10 for the sample discussed) is achieved at ∼1 MeV. (authors)

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.

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)

Method of neutralising the effects of electromagnetic radiation in a radiation detector and a radiation detector applying the procedure

International Nuclear Information System (INIS)

Gripentog, W.G.

1972-01-01

Circuitry is described by means of which radiation detectors of the Neher-White type, employing ionisation chambers can be unaffected by electromagnetic radiation which would otherwise cause inductive effects leading to erroneous signals. It is therefore unnecessary to use shielded cables for these instruments. (JIW)

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.

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)

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)

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.

A Prototype Large Area Detector Module for Muon Scattering Tomography

Energy Technology Data Exchange (ETDEWEB)

Steer, C.A.; Boakes, J.; Burns, J.; Snow, S.; Stapleton, M.; Thompson, L.F.; Quillin, S. [AWE Aldermaston, Reading, Berkshire, RG7 4PR (United Kingdom)

2015-07-01

Abstract-Shielded special nuclear materials (SNM) are of concern as some fissile isotopes have low gamma and neutron emission rates. These materials are also easily shielded to the point where their passive emissions are comparable to background. Consequently, shielded SNM is very challenging for passive radiation detection portals which scan cargo containers. One potential solution for this is to utilise the natural cosmic ray muon background and examine how these muons scatter from materials inside the container volume, terms; the muon scattering tomography (MST) technique measures the three-dimensional localised scattering at all points within a cargo container, providing a degree of material discrimination. There is the additional benefit that the MST signal increases with the presence of more high density shielding materials, in contrast to passive radiation detection. Simulations and calculations suggest that the effectiveness of the technique is sensitive to the tracking accuracy amongst other parameters, motivating the need to develop practical detector systems that are capable of tracking cosmic ray muons. To this end, we have constructed and tested a 2 m by 2 m demonstration module based on gaseous drift chambers and triggered by a large area scintillator-based detector, which is readout by wavelength shifting fibres. We discuss its design, construction, characterisation and operational challenges. (authors)

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

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.

Development of a shield based on Monte-Carlo studies for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine [Institut fuer Experimentalphysik, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

2013-07-01

COBRA is a next-generation experiment searching for neutrinoless double beta decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a decay energy of 2813.5 keV well above the highest naturally occurring gamma lines. The concept for a large scale set-up consists of an array of CdZnTe detectors with a total mass of 420 kg enriched in {sup 116}Cd up to 90 %. With a background rate in the order of 10{sup -3} counts/keV/kg/year, the experiment would be sensitive to a half-life larger than 10{sup 26} years, corresponding to a Majorana mass term m{sub ββ} smaller than 50 meV. To achieve the background level, an appropriate shield is necessary. The shield is developed based on Monte-Carlo simulations. For that, different materials and configurations are tested. In the talk the current status of the Monte-Carlo survey is presented and discussed.

Development of CANDLES low background HPGe detector and half-life measurement of 180Tam

Science.gov (United States)

Chan, W. M.; Kishimoto, T.; Umehara, S.; Matsuoka, K.; Suzuki, K.; Yoshida, S.; Nakajima, K.; Iida, T.; Fushimi, K.; Nomachi, M.; Ogawa, I.; Tamagawa, Y.; Hazama, R.; Takemoto, Y.; Nakatani, N.; Takihira, Y.; Tozawa, M.; Kakubata, H.; Trang, V. T. T.; Ohata, T.; Tetsuno, K.; Maeda, T.; Khai, B. T.; Li, X. L.; Batpurev, T.

2018-01-01

A low background HPGe detector system was developed at CANDLES Experimental Hall for multipurpose use. Various low background techniques were employed, including hermatic shield design, radon gas suppression, and background reduction analysis. A new pulse shape discrimination (PSD) method was specially created for coaxial Ge detector. Using this PSD method, microphonics noise and background event at low energy region less than 200 keV can be rejected effectively. Monte Carlo simulation by GEANT4 was performed to acquire the detection efficiency and study the interaction of gamma-rays with detector system. For rare decay measurement, the detector was utilized to detect the nature's most stable isomer tantalum-180m (180Tam) decay. Two phases of tantalum physics run were completed with total livetime of 358.2 days, which Phase II has upgraded shield configuration. The world most stringent half-life limit of 180Tam has been successfully achieved.

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)

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.

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)

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.

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.

Performance Evaluation of Spectroscopic Detectors for LEU Hold-up Measurements

Energy Technology Data Exchange (ETDEWEB)

Venkataraman, Ramkumar [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nutter, Greg [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-06

The hold-up measurement of low-enriched uranium materials may require use of alternate detector types relative to the measurement of highly enriched uranium. This is in part due to the difference in process scale (i.e., the components are generally larger for low-enriched uranium systems), but also because the characteristic gamma-ray lines from ²³⁵U used for assay of highly enriched uranium will be present at a much reduced intensity (on a per gram of uranium basis) at lower enrichments. Researchers at Oak Ridge National Laboratory examined the performance of several standard detector types, e.g., NaI(Tl), LaBr3(Ce), and HPGe, to select a suitable candidate for measuring and quantifying low-enriched uranium hold-up in process pipes and equipment at the Portsmouth gaseous diffusion plant. Detector characteristics, such as energy resolution (full width at half maximum) and net peak count rates at gamma ray energies spanning a range of 60–1332 keV, were measured for the above-mentioned detector types using the same sources and in the same geometry. Uranium enrichment standards (Certified Reference Material no. 969 and Certified Reference Material no. 146) were measured using each of the detector candidates in the same geometry. The net count rates recorded by each detector at 186 keV and 1,001 keV were plotted as a function of enrichment (atom percentage). Background measurements were made in unshielded and shielded configurations under both ambient and elevated conditions of ²³⁸U activity. The highly enriched uranium hold-up measurement campaign at the Portsmouth plant was performed on process equipment that had been cleaned out. Therefore, in most cases, the thickness of the uranium deposits was less than the “infinite thickness” for the 186 keV gamma rays to be completely self-attenuated. Because of this, in addition to measuring the 186 keV gamma, the 1,001 keV gamma ray from ^234mPa—a daughter of ²³⁸U in secular

Radiation-shielded double crystal X-ray monochromator for JET

International Nuclear Information System (INIS)

Barnsley, R.; Morsi, H.W.; Rupprecht, G.; Kaellne, E.

1989-01-01

A double crystal X-ray monochromator for absolute wavelength and intensity measurements with very effective shielding of its detector against neutrons and hard X-rays was brought into operation at JET. Fast wavelength scans were taken of impurity line radiation in the wavelength region from about 0.1 nm to 2.3 nm, and monochromatic as well as spectral line scans, for different operational modes of JET. (author) 5 refs., 4 figs

Fission signal detection using helium-4 gas fast neutron scintillation detectors

Energy Technology Data Exchange (ETDEWEB)

Lewis, J. M., E-mail: lewisj@ufl.edu; Kelley, R. P.; Jordan, K. A. [Nuclear Engineering Program, University of Florida, Gainesville, Florida 32611 (United States); Murer, D. [Arktis Radiation Detectors Ltd., 8045 Zurich (Switzerland)

2014-07-07

We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure {sup 4}He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the {sup 4}He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

Radiation shielding activities at IDOM

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Radiation shielding activities at IDOM

International Nuclear Information System (INIS)

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

2017-01-01

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

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

MARTA: A high-energy cosmic-ray detector concept with high-accuracy muon measurement

Energy Technology Data Exchange (ETDEWEB)

Abreu, P.; et al.

2017-12-20

A new concept for the direct measurement of muons in air showers is presented. The concept is based on resistive plate chambers (RPCs), which can directly measure muons with very good space and time resolution. The muon detector is shielded by placing it under another detector able to absorb and measure the electromagnetic component of the showers such as a water-Cherenkov detector, commonly used in air shower arrays. The combination of the two detectors in a single, compact detector unit provides a unique measurement that opens rich possibilities in the study of air showers.

40 K, 137 Cs and 232 Th activities in Brazilian milk samples measured by gamma ray spectrometry

International Nuclear Information System (INIS)

Melquiades, Fabio L.; Appoloni, Carlos R.

2000-01-01

This work deals with the measurement of radioactive activities in powdered milk, with high resolution gamma-ray spectrometry, using a HPGe detector coupled to a standard electronic nuclear chain and a multichannel card of 8192 channels. Preliminary measurements were accomplished to define the kind of the system shield, the geometry of the sample recipient, the size of the sampling and the self absorption correction. It was possible to measure the radionuclides 40 K, 137 Cs and 208 Tl, whose activities were calculated according to the International Atomic Energy Agency norms. The detector efficiency was measured employing calibrated samples, prepared with IAEA certificate standards mixed with powdered milk. Tukey's average comparison test was used to check the repeatability of the measurements and the absence of significant systematic deviation. (author)

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

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)

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...

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.

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

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.

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

Evaluation of an unshielded luminescence flow-through radio-HPLC detector for LC quality control and preparation of PET radiopharmaceuticals

International Nuclear Information System (INIS)

Thonon, David; Kaisin, Geoffroy; Henrottin, Jean; Aerts, Joël; Van Malderen, Hans; Luxen, André

2013-01-01

Radio-HPLC is an essential method to assess the purity of PET radiopharmaceuticals. The usual NaI scintillator radiodetector requires heavy, costly and cumbersome lead shielding. The luminescence LB 500 fLumo detector has been developed to tackle these drawbacks and achieve high sensitivity. The fLumo uses a photon counting detector combined with a flow-through cell modified with a solid melt-on scintillator only sensitive to the positron. This study demonstrates the usefulness of the fLumo for analysis and purification of PET radiopharmaceuticals. - Highlights: ► We evaluate a novel unshielded luminescence flow-through radio-HPLC detector (fLumo) which is only sensitive to the positron and insensitive to gamma rays for applications in PET radiopharmaceuticals analysis and preparation. ► The fLumo detector exhibits a low limit of detection as activities as low as 4 kBq are detected (HPLC and UPLC radiodetectors). ► The fLumo detector demonstrates excellent linearity (0.2 to 2500 MBq/ml, r 2 >0.995) and reproducibility. ► Thanks to its compactness and absence of shielding, the fLumo has been installed in a production shielded “hot” cell to detect radiocompounds during a semi-preparative HPLC purification. ► This work demonstrates the value of the fLumo luminescence flow-through radio-HPLC detector for applications in PET tracers radiochemistry

Monte Carlo Simulations of Ultra-High Energy Resolution Gamma Detectors for Nuclear Safeguards

International Nuclear Information System (INIS)

Robles, A.; Drury, O.B.; Friedrich, S.

2009-01-01

Ultra-high energy resolution superconducting gamma-ray detectors can improve the accuracy of non-destructive analysis for unknown radioactive materials. These detectors offer an order of magnitude improvement in resolution over conventional high purity germanium detectors. The increase in resolution reduces errors from line overlap and allows for the identification of weaker gamma-rays by increasing the magnitude of the peaks above the background. In order to optimize the detector geometry and to understand the spectral response function Geant4, a Monte Carlo simulation package coded in C++, was used to model the detectors. Using a 1 mm 3 Sn absorber and a monochromatic gamma source, different absorber geometries were tested. The simulation was expanded to include the Cu block behind the absorber and four layers of shielding required for detector operation at 0.1 K. The energy spectrum was modeled for an Am-241 and a Cs-137 source, including scattering events in the shielding, and the results were compared to experimental data. For both sources the main spectral features such as the photopeak, the Compton continuum, the escape x-rays and the backscatter peak were identified. Finally, the low energy response of a Pu-239 source was modeled to assess the feasibility of Pu-239 detection in spent fuel. This modeling of superconducting detectors can serve as a guide to optimize the configuration in future spectrometer designs.

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.

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.

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

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)

An InGrid based Low Energy X-ray Detector

CERN Document Server

Krieger, Christoph; Kaminski, Jochen; Lupberger, Michael; Vafeiadis, Theodoros

2014-01-01

An X-ray detector based on the combination of an integrated Micromegas stage with a pixel chip has been built in order to be installed at the CERN Axion Solar Telescope. Due to its high granularity and spatial resolution this detector allows for a topological background suppression along with a detection threshold below $1\\,\\text{keV}$. Tests at the CAST Detector Lab show the detector's ability to detect X-ray photons down to an energy as low as $277\\,\\text{eV}$. The first background data taken after the installation at the CAST experiment underline the detector's performance with an average background rate of $5\\times10^{-5}\\,/\\text{keV}/\\text{cm}^2/\\text{s}$ between 2 and $10\\,\\text{keV}$ when using a lead shielding.

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

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)

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

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

Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields

International Nuclear Information System (INIS)

Valdivia Valero, F. J.

2007-01-01

Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle θ 1 3, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs

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

Response function of semiconductor detectors, Ge and Si(Li); Funcao resposta de detectores semicondutores, Ge e Si(Li)

Energy Technology Data Exchange (ETDEWEB)

Zevallos Chavez, Juan Yury

2003-07-01

The Response Function (RF) for Ge and Si(Li) semiconductor detectors was obtained. The RF was calculated for five detectors, four Hp Ge with active volumes of 89 cm{sup 3} , 50 cm{sup 3} , 8 cm{sup 3} and 5 cm{sup 3}, and one Si(Li) with 0.143 cm{sup 3} of active volume. The interval of energy studied ranged from 6 keV up to 1.5 MeV. Two kinds of studies were done in this work. The first one was the RF dependence with the detection geometry. Here the calculation of the RF for a geometry named as simple and an extrapolation of that RF, were both done. The extrapolation process analyzed both, spectra obtained with a shielding geometry and spectra where the source-detector distance was modified. The second one was the RF dependence with the detection electronics. This study was done varying the shaping time of the pulse in the detection electronics. The purpose was to verify the effect of the ballistic deficit in the resolution of the detector. This effect was not observed. The RF components that describe the region of the total absorption of the energy of the incident photons, and the partial absorption of this energy, were both treated. In particular, empirical functions were proposed for the treatment of both, the multiple scattering originated in the detector (crystal), and the photon scattering originated in materials of the neighborhood of the crystal. Another study involving Monte Carlo simulations was also done in order to comprehend the photon scattering structures produced in an iron shield. A deconvolution method is suggested, for spectra related to scattered radiation in order to assess the dose delivered to the scatterer. (author)

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)

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.

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.)

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)

SuperCDMS Underground Detector Fabrication Facility

Energy Technology Data Exchange (ETDEWEB)

Platt, M.; Mahapatra, R.; Bunker, Raymond A.; Orrell, John L.

2018-03-01

The SuperCDMS SNOLAB dark matter experiment processes Ge and Si crystals into fully tested phonon and ionization detectors at surface fabrication and test facilities. If not mitigated, it is anticipated that trace-level production of radioisotopes in the crystals due to exposure to cosmic rays at (or above) sea level will result in the dominant source of background events in future dark matter searches using the current SuperCDMS detector technology. Fabrication and testing of detectors in underground facilities shielded from cosmic radiation is one way to directly reduce production of trace levels of radioisotopes, thereby improving experimental sensitivity for the discovery of dark matter beyond the level of the current experiment. In this report, we investigate the cost and feasibility to establish a complete detector fabrication processing chain in an underground location to mitigate cosmogenic activation of the Ge and Si detector substrates. For a specific and concrete evaluation, we explore options for such a facility located at SNOLAB, an underground laboratory in Sudbury, Canada hosting the current and future experimental phases of SuperCDMS.

Nuclear reactor ex-core startup neutron detector

International Nuclear Information System (INIS)

Wyvill, J.R.

1980-01-01

A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

A Cryogenic Detector Characterization Facility in the Shallow Underground Laboratory at the Technical University of Munich

Science.gov (United States)

Langenkämper, A.; Defay, X.; Ferreiro Iachellini, N.; Kinast, A.; Lanfranchi, J.-C.; Lindner, E.; Mancuso, M.; Mondragón, E.; Münster, A.; Ortmann, T.; Potzel, W.; Schönert, S.; Strauss, R.; Ulrich, A.; Wawoczny, S.; Willers, M.

2018-04-01

The Physics Department of the Technical University of Munich operates a shallow underground detector laboratory in Garching, Germany. It provides ˜ 160 {m^2} of laboratory space which is shielded from cosmic radiation by ˜ 6 m of gravel and soil, corresponding to a shielding of ˜ 15 {m.w.e.} . The laboratory also houses a cleanroom equipped with work- and wetbenches, a chemical fumehood as well as a spin-coater and a mask-aligner for photolithographic processing of semiconductor detectors. Furthermore, the shallow underground laboratory runs two high-purity germanium detector screening stations, a liquid argon cryostat and a ^3 He-^4 He dilution refrigerator with a base temperature of ≤ 12-14 mK . The infrastructure provided by the shallow laboratory is particularly relevant for the characterization of CaWO_4 target crystals for the CRESST-III experiment, detector fabrication and assembly for rare event searches. Future applications of the laboratory include detector development in the framework of coherent neutrino nucleus scattering experiments (ν -cleus) and studying its potential as a site to search for MeV-scale dark matter with gram-scale cryogenic detectors.

CDEX-1 1 kg point-contact germanium detector for low mass dark matter searches

International Nuclear Information System (INIS)

Kang Kejun; Yue Qian; Wu Yucheng

2013-01-01

The CDEX collaboration has been established for direct detection of light dark matter particles, using ultra-low energy threshold point-contact p-type germanium detectors, in China JinPing underground Laboratory (CJPL). The first 1 kg point-contact germanium detector with a sub-keV energy threshold has been tested in a passive shielding system located in CJPL. The outputs from both the point-contact P + electrode and the outside N + electrode make it possible to scan the lower energy range of less than 1 keV and at the same time to detect the higher energy range up to 3 MeV. The outputs from both P + and N + electrode may also provide a more powerful method for signal discrimination for dark matter experiment. Some key parameters, including energy resolution, dead time, decay times of internal X-rays, and system stability, have been tested and measured. The results show that the 1 kg point-contact germanium detector, together with its shielding system and electronics, can run smoothly with good performances. This detector system will be deployed for dark matter search experiments. (authors)

Feasibility study for use of a germanium detector in the LOFT gamma-ray densitometer

International Nuclear Information System (INIS)

Swierkowski, S.P.

1976-01-01

The primary aim of this study is to predict the performance of a gamma-ray densitometer system using computer modeling techniques. The system consists of a collimated 137 Cs source, a pipe containing a variable amount of water absorber, and a shielded and collimated germanium detector system. The gamma-ray energy spectrum (number of photon counts as a function of energy) has been computed for several sources at the detector. The response for combined sourceconfigurations has been obtained by linear superposition. The signal essentially consists of the counts in an energy window centered on the 137 Cs source at 662 keV that originate from this source. The noise is the background counts in the signal energy window that originate from 16 N scatter radiation and direct and shield tank activation gammas. The detector signal has been computed for 0, 50, and 100 percent water in the pipe

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.)

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)

Silicon for ultra-low-level detectors and sup 32 Si

Energy Technology Data Exchange (ETDEWEB)

Plaga, R. (Max Planck Inst. fuer Kernphysik, Heidelberg (Germany))

1991-11-15

A recent dark matter experiment using a silicon diode detector confirms that the decay of {sup 32}Si is a dangerous background in ultra-low-level experiments using silicon as detector material or shielding. In this Letter we study the mechanism of how {sup 32}Si enters commercially available silicon. Ways to avoid this contamination are pointed out. Limits on the {sup 32}Si content of silicon from measurements with miniaturized low-level proportional counters are also given. (orig.).

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

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.

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.

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.)

Study of Neutrino-Induced Neutrons in Dark Matter Detectors for Supernova Burst Neutrinos

Science.gov (United States)

Kwan, Newton; Scholberg, Kate

2017-09-01

When supernova burst neutrinos (1-50 MeV) pass through the Earth, they occasionally interact with the passive shielding surrounding dark matter detectors. When the neutrinos interact, one or two roughly 2 MeV neutrons are scattered isotropically and uniformly, often leaving undetected. Occasionally, these neutrino-induced neutrons (NINs) interact with the detector and leave a background signal similar to a WIMP. The purpose of this study is to understand the effects of NINs on active dark matter detectors during a supernova burst.

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)

Methods for radiation detection and characterization using a multiple detector probe

Science.gov (United States)

Akers, Douglas William; Roybal, Lyle Gene

2014-11-04

Apparatuses, methods, and systems relating to radiological characterization of environments are disclosed. Multi-detector probes with a plurality of detectors in a common housing may be used to substantially concurrently detect a plurality of different radiation activities and types. Multiple multi-detector probes may be used in a down-hole environment to substantially concurrently detect radioactive activity and contents of a buried waste container. Software may process, analyze, and integrate the data from the different multi-detector probes and the different detector types therein to provide source location and integrated analysis as to the source types and activity in the measured environment. Further, the integrated data may be used to compensate for differential density effects and the effects of radiation shielding materials within the volume being measured.

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.

Experiments on iron shield transmission of quasi-monoenergetic neutrons generated by 43- and 68-MeV protons via the 7Li(p,n) reaction

International Nuclear Information System (INIS)

Nakashima, Hiroshi; Tanaka, Shun-ichi; Nakao, Noriaki

1996-03-01

In order to provide benchmark data of neutrons transmitted through iron shields in the intermediate-energy region, spatial distributions of neutron energy spectra and reaction rates behind and inside the iron shields of thickness up to 130 cm were measured for 43- and 68-MeVp- 7 Li neutrons using a quasi-monoenergetic neutron beam source at the 90-MV AVF cyclotron facility of the TLARA facility in JAERI. The measured data by five kinds of detectors: the BC501A detector, the Bonner ball counter, 238 U and 232 Th fission counters, 7 LiF and nat LiF TLDs and solid state nuclear track detector, are numerically provided in this report in the energy region between 10 -4 eV and the energy of peak neutrons generated by the 7 Li(p,n) reaction. (author)

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....

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

International Nuclear Information System (INIS)

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

2006-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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

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,

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.

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.

Experimental investigation of clay fly ash bricks for gamma-ray shielding

Energy Technology Data Exchange (ETDEWEB)

Mann Harjinder Singh; Mudahar, Gumel Singh [Dept. of Physics, Punjabi University, Patiala (India); Brar, Gurdarshan Singh [Dept. of Higher Education, Additional Project Director, Chandigarh (India); Mann, Kulwinder Singh [Dept. of Applied Sciences, I.K. Gujral Punjab Technical University, Jalandhar (India)

2016-10-15

This study aims to determine the effect of fly ash with a high replacing ratio of clay on the radiation shielding properties of bricks. Some interaction parameters (mass attenuation coefficients, half value layer, effective atomic number, effective electron density, and absorption efficiency) of clay fly ash bricks were measured with a NaI(Tl) detector at 661.6 keV, 1,173.2 keV, and 1,332.5 keV. For the investigation of their shielding behavior, fly ash bricks were molded using an admixture to clay. A narrow beam transmission geometry condition was used for the measurements. The measured values of these parameters were found in good agreement with the theoretical calculations. The elemental compositions of the clay fly ash bricks were analyzed by using an energy dispersive X-ray fluorescence spectrometer. At selected energies the values of the effective atomic numbers and effective electron densities showed a very modest variation with the composition of the fly ash. This seems to be due to the similarity of their elemental compositions. The obtained results were also compared with concrete, in order to study the effect of fly ash content on the radiation shielding properties of clay fly ash bricks. The clay fly ash bricks showed good shielding properties for moderate energy gamma rays. Therefore, these bricks are feasible and eco-friendly compared with traditional clay bricks used for construction.

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)

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.)

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.

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

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

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.)

The Muon system of the run II D0 detector

Energy Technology Data Exchange (ETDEWEB)

Abazov, V.M.; Acharya, B.S.; Alexeev, G.D.; Alkhazov, G.; Anosov, V.A.; Baldin, B.; Banerjee, S.; Bardon, O.; Bartlett, J.F.; Baturitsky, M.A.; Beutel, D.; Bezzubov,; Bodyagin, V.; Butler, J.M.; Cease, H.; Chi, E.; Denisov, D.; Denisov, S.P.; Diehl, H.T.; Doulas, S.; Dugad, S.R.; /Beijing, Inst. High Energy Phys. /Charles U. /Prague, Tech.

2005-03-01

The authors describe the design, construction and performance of the upgraded D0 muon system for Run II of the Fermilab Tevatron collider. Significant improvements have been made to the major subsystems of the D0 muon detector: trigger scintillation counters, tracking detectors, and electronics. The Run II central muon detector has a new scintillation counter system inside the iron toroid and an improved scintillation counter system outside the iron toroid. In the forward region, new scintillation counter and tracking systems have been installed. Extensive shielding has been added in the forward region. A large fraction of the muon system electronics is also new.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Study of the background neutron and gamma components of the ββ(0ν) decay in the NEMO2 prototype detector. Consequences for the NEMO3 detector

International Nuclear Information System (INIS)

Marquet, Christine

1999-01-01

Neutrinoless double beta decay ββ(0ν) is a test of physics beyond the Standard Model by involving the existence of a massive Majorana neutrino (ν = ν-bar). To try to observe such a process with a sensitivity of 0.1 eV on the neutrino effective mass ( ν >), NEMO collaboration build the NEMO3 detector, able to measure half-lives greater than 10 24 years, corresponding to a few detected events per year. For that, it is necessary to know and master all background sources. This work was first dedicated to the study of external (to the double beta source) background with crossing electrons recorded with NEMO2 prototype detector and then to the simulation of this background in NEMO3 detector. Comparison between NEMO2 data and results of gamma and neutron simulations for different shieldings, with and without neutron source, has allowed to determine background contributions of radon, thoron, 208 Tl contaminations in materials, photon flux produced in laboratory and neutrons. This study, which has required improvements in the MICAP neutron simulation code by developing a photon generator, proved that radiative capture of fast neutrons thermalized in the detector was the source of events in the energy domain of the ββ(0ν) signal. In order to reach the required sensitivity on ν > mass, it has been shown that both a neutron shielding and magnetic field are necessary for NEMO3 detector. (author) [fr

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.

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.

Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

International Nuclear Information System (INIS)

Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

1994-01-01

Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

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.

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

Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

International Nuclear Information System (INIS)

Heidrich, Nadine

2014-11-01

The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10 -3 counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm 3 CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10 -6 counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a contribution of 26

Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine

2014-11-15

The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10{sup -3} counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm{sup 3} CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10{sup -6} counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a

Evaluation of K x-ray escape and crosstalk in CdTe detectors and multi-channel detectors

International Nuclear Information System (INIS)

Ohtsuchi, Tetsuro; Ohmori, Koichi; Tsutsui, Hiroshi; Baba, Sueki

1995-01-01

The simple structure of CdTe semiconductor detectors facilitates their downsizing, and their possible application to radiographic sensors has been studied. The escape of K X-rays from these detectors increases with reduction of their dimensions and affects the measurements of X- and gamma-ray spectra. K X-rays also produce crosstalk in multi-channel detectors with adjacent channels. Therefore, K X-rays which escape from the detector elements degrade both the precision of energy spectra and spatial resolution. The ratios of escape peak integrated counts to total photon counts for various sizes of CdTe single detectors were calculated for gamma rays using the Monte Carlo method. Also, escape and crosstalk ratios were simulated for the CdTe multi-channel detectors. The theoretical results were tested experimentally for 59.54-keV gamma rays from a 241 Am radioactive source. Results showed that escape ratios for single detectors were strongly dependent on element size and thickness. The escape and crosstalk ratios increased with closer channel pitch. The calculated results showed a good agreement with the experimental data. The calculations made it clear that K X-rays which escaped to neighboring channels induced crosstalk more frequently at smaller channel pitch in multichannel detectors. A radiation shielding grid which blocked incident photons between the boundary channels was also tested by experiment and by calculation. It was effective in reducing the probability of escape and crosstalk

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Direct measurement of homogeneously distributed radioactive air contamination with germanium detectors

International Nuclear Information System (INIS)

Sowa, W.

1990-01-01

Air contamination by γ emitting radionuclides was measured with a vertically arranged germanium detector, laterally shielded by a lead ring, and calibration factors and detection limits of a number of fission products determined. The possibility of measuring simultaneously existing air and soil contamination by measurements with and without lead shield is described. The change of detection limit of air contamination is presented for different soil contamination levels by the same radionuclide. Calibration factors are given to determine the dose rate on the ground due to air contamination by different radionuclides. (author)

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)

Female gonadal shielding with automatic exposure control increases radiation risks

Energy Technology Data Exchange (ETDEWEB)

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

2018-02-15

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

Female gonadal shielding with automatic exposure control increases radiation risks

International Nuclear Information System (INIS)

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

2018-01-01

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

Female gonadal shielding with automatic exposure control increases radiation risks.

Science.gov (United States)

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

2018-02-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

The shielding calculation for the CN guide shielding assembly in HANARO

International Nuclear Information System (INIS)

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

2006-01-01

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

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.

Detection and Imaging of High-Z Materials with a Muon Tomography Station Using GEM Detectors

CERN Document Server

Gnanvo, K; Bittner, W; Costa, F; Grasso, L; Hohlmann, M; Locke, J B; Martoiu, S; Muller, H; Staib, M; Tarazona, A; Toledo, J

2010-01-01

Muon tomography based on the measurement of multiple scattering of atmospheric cosmic ray muons is a promising technique for detecting and imaging heavily shielded high-Z nuclear materials such as enriched uranium. This technique could complement standard radiation detection portals currently deployed at international borders and ports, which are not very sensitive to heavily shielded nuclear materials. We image small targets in 3D using $2\\times 2 \\times 2$ mm^3 voxels with a minimal muon tomography station prototype that tracks muons with Gas Electron Multiplier (GEM) detectors read out in 2D with x-y microstrips of 400 micron pitch. With preliminary electronics, the GEM detectors achieve a spatial resolution of 130 microns in both dimensions. With the next GEM-based prototype station we plan to probe an active volume of ~27 liters. We present first results on reading out all 1536 microstrips of a $30 \\times 30$ cm^2 GEM detector for the next muon tomography prototype with final frontend electronics and DAQ...

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.

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

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

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

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.

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.)

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).

Experiments on iron shield transmission of quasi-monoenergetic neutrons generated by 43- and 68-MeV protons via the {sup 7}Li(p,n) reaction

Energy Technology Data Exchange (ETDEWEB)

Nakashima, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tanaka, Shun-ichi; Nakao, Noriaki [and others

1996-03-01

In order to provide benchmark data of neutrons transmitted through iron shields in the intermediate-energy region, spatial distributions of neutron energy spectra and reaction rates behind and inside the iron shields of thickness up to 130 cm were measured for 43- and 68-MeVp-{sup 7}Li neutrons using a quasi-monoenergetic neutron beam source at the 90-MV AVF cyclotron facility of the TLARA facility in JAERI. The measured data by five kinds of detectors: the BC501A detector, the Bonner ball counter, {sup 238}U and {sup 232}Th fission counters, {sup 7}LiF and {sup nat}LiF TLDs and solid state nuclear track detector, are numerically provided in this report in the energy region between 10{sup -4} eV and the energy of peak neutrons generated by the {sup 7}Li(p,n) reaction. (author).

A microwave paraphoton and axion detection experiment with 300 dB electromagnetic shielding at 3 GHz

CERN Document Server

Betz, M

2012-01-01

For the microwave equivalent of “light shining through the wall” (LSW) experiments, a sensitive microwave detector and very high electromagnetic shielding is required. The screening attenuation between the axion generating cavity and the nearby detection cavity should be greater than 300 dB, in order to improve over presently existing exclusion limits. To achieve these goals in practice, a “box in a box” concept was utilized for shielding the detection cavity, while a vector signal analyzer was used as a microwave receiver with a very narrow resolution bandwidth in the order of a few micro-Hz. This contribution will present the experimental layout and the results to date.

A novel optically transparent RF shielding for fully integrated PET/MRI systems

Science.gov (United States)

Parl, C.; Kolb, A.; Schmid, A. M.; Wehrl, H. F.; Disselhorst, J. A.; Soubiran, P. D.; Stricker-Shaver, D.; Pichler, B. J.

2017-09-01

Preclinical imaging benefits from simultaneous acquisition of high-resolution anatomical and molecular data. Additionally, PET/MRI systems can provide functional PET and functional MRI data. To optimize PET sensitivity, we propose a system design that fully integrates the MRI coil into the PET system. This allows positioning the scintillators near the object but requires an optimized design of the MRI coil and PET detector. It further requires a new approach in realizing the radiofrequency (RF) shielding. Thus, we propose the use of an optically transparent RF shielding material between the PET scintillator and the light sensor, suppressing the interference between both systems. We evaluated two conductive foils (ITO, 9900) and a wire mesh. The PET performance was tested on a dual-layer scintillator consisting of 12  ×  12 LSO matrices, shifted by half a pitch. The pixel size was 0.9  ×  0.9 mm2 the lengths were 10.0 mm and 5.0 mm, respectively. For a light sensor, we used a 4  ×  4 SiPM array. The RF attenuation was measured from 320 kHz to 420 MHz using two pick-up coils. MRI-compatibility and shielding effect of the materials were evaluated with an MRI system. The average FWHM energy resolution at 511 keV of all 144 crystals of the layer next to the SiPM was deteriorated from 15.73  ±  0.24% to 16.32  ±  0.13%, 16.60  ±  0.25%, and 19.16  ±  0.21% by the ITO foil, 9900 foil, mesh material, respectively. The average peak-to-valley ratio of the PET detector changed from 5.77  ±  0.29 to 4.50  ±  0.39, 4.78  ±  0.48, 3.62  ±  0.16, respectively. The ITO, 9900, mesh attenuated the scintillation light by 11.3  ±  1.6%, 11.0  ±  1.8%, 54.3  ±  0.4%, respectively. To attenuate the RF from 20 MHz to 200 MHz, mesh performed better than copper. The results show that an RF shielding material that is sufficiently transparent for

A more rugged ZnS(Ag) alpha scintillation detector

International Nuclear Information System (INIS)

McElhaney, S.A.; Ramsey, J.A.; Bauer, M.L.; Chiles, M.M.

1990-01-01

Conventional alpha scintillation detectors comprise a phosphor-coated light-pipe covered by a thin aluminized Mylar layer. This opaque radiation entrance window serves as a shield against ambient light entering the detector with minimum alpha attenuation. Unfortunately, Mylar is extremely fragile and easily punctured or torn by sticks, stones, and screws encountered during regular radiation surveys. The authors have been developing an alpha scintillation detector more rugged and durable than conventional models. This paper presents the scintillator assembly, which consists of a mixture of silver-activated zinc sulfide [ZnS(Ag)] and clear epoxy. The ZnS(Ag) scintillation powder is mixed with a low-viscosity, optically transparent epoxy and poured into a glass-smooth mold of desired shape and size

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.)

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

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

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.

Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

International Nuclear Information System (INIS)

Sekine, T.; Baba, H.

1978-01-01

Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

Report on the feasibility of the in situ radiometric determination of uranium grade in Witwatersrand gold and uranium mines

International Nuclear Information System (INIS)

Smit, C.J.B.; Wesolinski, E.S.; Corner, B.

1982-08-01

The chip-sampling technique currently employed by the South African gold and uranium-mining industry, for the prediction of face grade, has several drawbacks, namely: 1) it is labour-intensive; 2) sample volumes are often unrepresentative and prone to human error; and 3) the uranium mineralisation may be very erratic along the reef. In situ radiometric assaying for uranium along the reef, on the other hand, is a rapid, essentially one-man operation, enabling a much larger and hence a more representative sample volume to be measured. The high radiometric background inherent in any uranium mine necessitates some form of high-density shielding in order to facilitate quantitative in situ assaying. This report, therefore, briefly outlines the origin, nature, detection and shielding of gamma rays. Results obtained with a frontally shielded total-count instrument showed that radiometric estimates of uranium grade are comparable to those obtained by batch mining and can be used for the prediction of face grades, provided that the ore is in radiometric equilibrium and that thorium and potassium are either not present, or vary sympathetically with the uranium grade. Spectral analysis showed, however, that these circumstances will also permit the use of a collimated (side-shielded) detector of acceptable weight, provided that only the low-energy portion of the spectrum is measured. The advantages of a collimated detector over a frontally shielded detector are also noteworthy, viz.: 1) only one reading is taken per sample point rather than two, as is the case with the frontally shielded system, thus improving counting statistics; and 2) the shielding is permanently fixed to the detector. Comprehensive design considerations for a compact, portable instrument are suggested and methods for determining background radiation as applicable to a collimated detector are described

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).

Adiabatic Demagnetization Refrigerator Field Mapping and Shielding Models for a 70 mK Superconducting Transition Edge Sensor Array and Associated Electronics

Science.gov (United States)

Ladner, D. R.; Martinez-Galarce, D. S.; McCammon, D.

2006-04-01

An X-ray detection instrument to be flown on a sounding rocket experiment (the Advanced Technology Solar Spectroscopic Imager - ATSSI) for solar physics observations is being developed by the Lockheed Martin Solar and Astrophysics Laboratory (LMSAL). The detector is a novel class of microcalorimeter, a superconducting Transition-Edge Sensor (TES), that coupled with associated SQUID and feedback electronics requires high temperature stability at ~70 mK to resolve the energy of absorbed X-ray photons emitted from the solar corona. The cooling system incorporates an existing Adiabatic Demagnetization Refrigerator (ADR) developed at the University of Wisconsin (UW), which was previously flown to study the diffuse cosmic X-ray background. The Si thermistor detectors for that project required 130 K shielded JFET electronic components that are much less sensitive to the external field of the ADR solenoid than are the 1st (~70 mK) and 2nd (~2 K) SQUID stages used with TESs for solar observations. Modification of the Wisconsin ADR design, including TES focal plane and electronics re-positioning, therefore requires a tradeoff between the existing ADR solenoid nulling coil geometry and a low mass passive solenoid shield, while preserving the vibration isolation features of the existing design. We have developed models to accurately compute the magnetic field with and without shielding or nulling coils at critical locations to guide the re-design of the detector subsystem. The models and their application are described.

The HERMES dual-radiator ring imaging Cherenkov detector

CERN Document Server

Akopov, N; Bailey, K; Bernreuther, S; Bianchi, N; Capitani, G P; Carter, P; Cisbani, E; De Leo, R; De Sanctis, E; De Schepper, D; Dzhordzhadze, V; Filippone, B W; Frullani, S; Garibaldi, F; Hansen, J O; Hommez, B; Iodice, M; Jackson, H E; Jung, P; Kaiser, R; Kanesaka, J; Kowalczyk, R; Lagamba, L; Maas, A; Muccifora, V; Nappi, E; Negodaeva, K; Nowak, Wolf-Dieter; O'Connor, T; O'Neill, T G; Potterveld, D H; Ryckbosch, D; Sakemi, Y; Sato, F; Schwind, A; Shibata, T A; Suetsugu, K; Thomas, E; Tytgat, M; Urciuoli, G M; Van De Kerckhove, K; Van De Vyver, R; Yoneyama, S; Zhang, L F; Zohrabyan, H G

2002-01-01

The construction and use of a dual radiator Ring Imaging Cherenkov (RICH) detector is described. This instrument was developed for the HERMES experiment at DESY which emphasises measurements of semi-inclusive deep-inelastic scattering. It provides particle identification for pions, kaons, and protons in the momentum range from 2 to 15 GeV, which is essential to these studies. The instrument uses two radiators, C sub 4 F sub 1 sub 0 , a heavy fluorocarbon gas, and a wall of silica aerogel tiles. The use of aerogel in a RICH detector has only recently become possible with the development of clear, large, homogeneous and hydrophobic aerogel. A lightweight mirror was constructed using a newly perfected technique to make resin-coated carbon-fiber surfaces of optical quality. The photon detector consists of 1934 photomultiplier tubes (PMT) for each detector half, held in a soft steel matrix to provide shielding against the residual field of the main spectrometer magnet.

Response function of semiconductor detectors, Ge and Si(Li)

International Nuclear Information System (INIS)

Zevallos Chavez, Juan Yury

2003-01-01

The Response Function (RF) for Ge and Si(Li) semiconductor detectors was obtained. The RF was calculated for five detectors, four Hp Ge with active volumes of 89 cm 3 , 50 cm 3 , 8 cm 3 and 5 cm 3 , and one Si(Li) with 0.143 cm 3 of active volume. The interval of energy studied ranged from 6 keV up to 1.5 MeV. Two kinds of studies were done in this work. The first one was the RF dependence with the detection geometry. Here the calculation of the RF for a geometry named as simple and an extrapolation of that RF, were both done. The extrapolation process analyzed both, spectra obtained with a shielding geometry and spectra where the source-detector distance was modified. The second one was the RF dependence with the detection electronics. This study was done varying the shaping time of the pulse in the detection electronics. The purpose was to verify the effect of the ballistic deficit in the resolution of the detector. This effect was not observed. The RF components that describe the region of the total absorption of the energy of the incident photons, and the partial absorption of this energy, were both treated. In particular, empirical functions were proposed for the treatment of both, the multiple scattering originated in the detector (crystal), and the photon scattering originated in materials of the neighborhood of the crystal. Another study involving Monte Carlo simulations was also done in order to comprehend the photon scattering structures produced in an iron shield. A deconvolution method is suggested, for spectra related to scattered radiation in order to assess the dose delivered to the scatterer. (author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

2011-01-01

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

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.

Functional Requirements on the Design of the Detectors and the Interaction Region of an e+e- Linear Collider with a Push-Pull Arrangement of Detectors

International Nuclear Information System (INIS)

Markiewicz, T.

2009-01-01

The Interaction Region of the International Linear Collider is based on two experimental detectors working in a push-pull mode. A time efficient implementation of this model sets specific requirements and challenges for many detector and machine systems, in particular the IR magnets, the cryogenics and the alignment system, the beamline shielding, the detector design and the overall integration. This paper attempts to separate the functional requirements of a push pull interaction region and machine detector interface from any particular conceptual or technical solution that might have been proposed to date by either the ILC Beam Delivery Group or any of the three detector concepts. As such, we hope that it provides a set of ground rules for interpreting and evaluating the MDI parts of the proposed detector concept's Letters of Intent, due March 2009. The authors of the present paper are the leaders of the IR Integration Working Group within Global Design Effort Beam Delivery System and the representatives from each detector concept submitting the Letters Of Intent.

Automation of the Characterization of High Purity Germanium Detectors

Science.gov (United States)

Dugger, Charles ``Chip''

2014-09-01

Neutrinoless double beta decay is a rare hypothesized process that may yield valuable insight into the fundamental properties of the neutrino. Currently there are several experiments trying to observe this process, including the Majorana DEMONSTRAOR experiment, which uses high purity germanium (HPGe) detectors to generate and search for these events. Because the event happens internally, it is essential to have the lowest background possible. This is done through passive detector shielding, as well as event discrimination techniques that distinguish between multi-site events characteristic of gamma-radiation, and single-site events characteristic of neutrinoless double beta decay. Before fielding such an experiment, the radiation response of the detectors must be characterized. A robotic arm is being tested for future calibration of HPGe detectors. The arm will hold a source at locations relative to the crystal while data is acquired. Several radioactive sources of varying energy levels will be used to determine the characteristics of the crystal. In this poster, I will present our work with the robot, as well as the characterization of data we took with an underground HPGe detector at the WIPP facility in Carlsbad, NM (2013). Neutrinoless double beta decay is a rare hypothesized process that may yield valuable insight into the fundamental properties of the neutrino. Currently there are several experiments trying to observe this process, including the Majorana DEMONSTRAOR experiment, which uses high purity germanium (HPGe) detectors to generate and search for these events. Because the event happens internally, it is essential to have the lowest background possible. This is done through passive detector shielding, as well as event discrimination techniques that distinguish between multi-site events characteristic of gamma-radiation, and single-site events characteristic of neutrinoless double beta decay. Before fielding such an experiment, the radiation response of

The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)

Radiation Build-Up In Shielding Of Low Activity High Energia Gamma Source

International Nuclear Information System (INIS)

Helfi-Yuliati; Mukhlis-Akhadi

2003-01-01

Research to observe radiation build-up factor (b) in aluminium (Al), iron (Fe) and lead (Pb) for shielding of gamma radiation of high energy from 137 cs (E γ : 662 keV) source and 60 Co (E γ : 1332 keV) of low activity sources has been carried out. Al with Z =13 represent metal of low atomic number, Fe with Z =26 represent metal of medium atomic number, and Pb with Z = 82 represent metal of high atomic number. Low activity source in this research is source which if its dose rate decrease to 3 % of its initial dose rate became safe for the workers. Research was conducted by counting of radiation intensity behind shielding with its thickness vary from 1 to 5 times of half value thickness (HVT). NaI(TI) detector which connected to multi channel analyzer (MCA) was used for the counting. Calculation result show that all of b value are close to 1 (b ∼ 1) for all kinds of metals. No radiation build-up factor is required in estimating the shielding thickness from several kinds of metals for low activity of high energy gamma source. (author)

REACTOR SHIELD

Science.gov (United States)

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

1959-02-17

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

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

Science.gov (United States)

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

2016-09-08

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

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)

A method of detector correction for cosmic ray muon radiography

International Nuclear Information System (INIS)

Liu Yuanyuan; Zhao Ziran; Chen Zhiqiang; Zhang Li; Wang Zhentian

2008-01-01

Cosmic ray muon radiography which has good penetrability and sensitivity to high-Z materials is an effective way for detecting shielded nuclear materials. The problem of data correction is one of the key points of muon radiography technique. Because of the influence of environmental background, environmental yawp and error of detectors, the raw data can not be used directly. If we used the raw data as the usable data to reconstruct without any corrections, it would turn up terrible artifacts. Based on the characteristics of the muon radiography system, aimed at the error of detectors, this paper proposes a method of detector correction. The simulation experiments demonstrate that this method can effectively correct the error produced by detectors. Therefore, we can say that it does a further step to let the technique of cosmic muon radiography into out real life. (authors)

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)

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

Evaluation of the shielding integrity of end-shields in PHWR type NPPs

International Nuclear Information System (INIS)

Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

1996-01-01

In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

Characterization of Compton-suppressed TIGRESS detectors for high energy gamma-rays

International Nuclear Information System (INIS)

Kshetri, R.; Andreoiu, C.; Cross, D.S.; Galinski, N.; Ball, G.C.; Djongolov, M.; Garnsworthy, A.B.; Hackman, G.; Orce, J.N.; Pearson, C.; Triambak, S.; Williams, S.J.; Drake, T.; Smalley, D.; Svensson, C.E.

2009-01-01

The TRIUMF-ISAC Gamma-Ray Escape- Suppressed Spectrometer (TIGRESS) will consist of 12 large-volume, 32-fold segmented HPGe clover detectors. Each detector is shielded by a 20-fold segmented Compton suppression shield. For performing discrete gamma-ray spectroscopy of light mass nuclei with TIGRESS, we need information about full energy peak efficiency, resolution and lineshape of full energy peaks for high energy gamma-rays. However, suitable radioactive sources having decay gamma-rays of energies greater than ∼ 3.5 MeV are not easily available. So the characteristics of gamma spectrometers at energies higher than 3.5 MeV are usually determined from simulation data. Predictions from GEANT4 simulations (experimentally validated from 0.3 to 3 MeV) indicate that TIGRESS will be capable for single 10 MeV gamma-rays of absolute detection efficiency of 1.5% for backward configuration of the array. It has been observed experimentally that simulation results work well up to certain energies and might deviate at higher energies. So, it is essential to check the validity of simulation results for energies above 3.3 MeV. We have investigated the high energy performance of seven TIGRESS detectors up to 8 MeV

Composites with carbon nanotube for radiation shielding application

International Nuclear Information System (INIS)

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

2017-01-01

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

Composites with carbon nanotube for radiation shielding application

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

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.

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

Using the IEC Standard to Describe Low-Background Detectors-What Can You Expect?

International Nuclear Information System (INIS)

Ronald M. Keyser; Sanford Wagner

1998-01-01

The new International Electrotechnical Commission (IEC) standard for describing the background makes the specification of the background in a high-purity germanium (HPGe) detector simple, unambiguous, and related to how the detector will be used. Users and manufacturers will finally be speaking the same language on this subject. Because this standard extends the specification of the performance of an HPGe detector, there is little history available for comparison and thus no means of determining a ''good'' value. To develop a history, the background spectrum for 500 low-background HPGe ORTEC detectors were all counted in similar low-background shields. These detectors were in a variety of mechanical cryostat and endcap configurations. The continuum background is a function of energy and detector size/configuration. The peak area for the peak energies listed in the standard is a function of detector size and configuration. The results thus give practical guidance for obtaining the most appropriate low-background detector for a specific measurement problem

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

High-performance heavy concrete as a multi-purpose shield

International Nuclear Information System (INIS)

Mortazavi, S. M. J.; Mosleh-Shirazi, M. A.; Roshan-Shomal, P.; Raadpey, N.; Baradaran-Ghahfarokhi, M.

2010-01-01

Concrete has long been used as a shield against high-energy photons and neutrons. In this study, colemanite and galena minerals (CoGa) were used for the production of an economical high-performance heavy concrete. To measure the gamma radiation attenuation of the CoGa concrete samples, they were exposed to a narrow beam of gamma rays emitted from a 60 Co radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. The compression strengths of both types of concrete mixes (CoGa and reference concrete) were investigated. The range of the densities of the heavy concrete samples was 4100-4650 kg m -3 , whereas it was 2300-2600 kg m -3 in the ordinary concrete reference samples. The half-value layer of the CoGa concrete samples for 60 Co gamma rays was 2.49 cm; much less than that of ordinary concrete (6.0 cm). Moreover, CoGa concrete samples had a 10% greater neutron absorption compared with reference concrete. (authors)

Aerodynamic sampling for landmine trace detection

Science.gov (United States)

Settles, Gary S.; Kester, Douglas A.

2001-10-01

Electronic noses and similar sensors show promise for detecting buried landmines through the explosive trace signals they emit. A key step in this detection is the sampler or sniffer, which acquires the airborne trace signal and presents it to the detector. Practicality demands no physical contact with the ground. Further, both airborne particulates and molecular traces must be sampled. Given a complicated minefield terrain and microclimate, this becomes a daunting chore. Our prior research on canine olfactory aerodynamics revealed several ways that evolution has dealt with such problems: 1) proximity of the sniffer to the scent source is important, 2) avoid exhaling back into the scent source, 3) use an aerodynamic collar on the sniffer inlet, 4) use auxiliary airjets to stir up surface particles, and 5) manage the 'impedance mismatch' between sniffer and sensor airflows carefully. Unfortunately, even basic data on aerodynamic sniffer performance as a function of inlet-tube and scent-source diameters, standoff distance, etc., have not been previously obtained. A laboratory-prototype sniffer was thus developed to provide guidance for landmine trace detectors. Initial experiments with this device are the subject of this paper. For example, a spike in the trace signal is observed upon starting the sniffer airflow, apparently due to rapid depletion of the available signal-laden air. Further, shielding the sniffer from disruptive ambient airflows arises as a key issue in sampling efficiency.

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.)

Testicular shield for para-aortic radiotherapy and estimation of gonad doses.

Science.gov (United States)

Ravichandran, R; Binukumar, J P; Kannadhasan, S; Shariff, M H; Ghamrawy, Kamal El

2008-10-01

For radiotherapy of para-aortic and abdominal regions in male patients, gonads are to be protected to receive less than 2% of the prescribed dose. A testicular shield was fabricated for abdominal radiotherapy with 15 MV X-rays ((Clinac 2300 CD, Varian AG) with low melting point alloy (Cerroband). The dimensions of the testicular shield were 6.5 cm diameter and 3.5 cm depth with 1.5 cm wall thickness. During treatment, this shield was held in position by a rectangular sponge and Styrofoam support. Phantom measurement was carried out with a humanoid phantom and a 0.6 cc ion chamber. The mean energy of the scattered photon was calculated for single scattering at selected distances from the beam edge and with different field dimensions. One patient received radiotherapy with an inverted Y field and gonad doses were estimated using calibrated thermo-luminescent detector (TLD) chips. Measured doses with the ion chamber were 7.1 and 3.5% of the mid-plane doses without a shield at 3 and 7.5 cm off-field respectively. These values decreased to 4.6 and 1.7% with the bottom shield alone, and to 1.7 and 0.8% with both bottom and top shields covering the ion chamber. The measured doses at the gonads during the patient's treatment were 0.5-0.92% for the AP field (0.74 +/- 0.17%, n = 5) and 0.5-1.2% for the PA field (0.88 +/- 0.24%, n = 5). The dose received by the testis for the full course of treatment was 32 cGy (0.8%) for a total mid-plane dose of 40 Gy. The first-scatter energy estimated at the gonads is around 1.14 MeV for a primary beam of 15 MV for a long axis dimension of 37 cm of primary field. During the patient's treatment, the estimated absorbed doses at the gonads were comparable with reported values in similar treatments. The testicular shield reported in this study is of light weight and could be used conveniently in treatments of abdominal fields.

Testicular shield for para-aortic radiotherapy and estimation of gonad doses

Directory of Open Access Journals (Sweden)

Ravichandran R

2008-01-01

Full Text Available For radiotherapy of para-aortic and abdominal regions in male patients, gonads are to be protected to receive less than 2% of the prescribed dose. A testicular shield was fabricated for abdominal radiotherapy with 15 MV X-rays ((Clinac 2300 CD, Varian AG with low melting point alloy (Cerroband. The dimensions of the testicular shield were 6.5 cm diameter and 3.5 cm depth with 1.5 cm wall thickness. During treatment, this shield was held in position by a rectangular sponge and Styrofoam support. Phantom measurement was carried out with a humanoid phantom and a 0.6 cc ion chamber. The mean energy of the scattered photon was calculated for single scattering at selected distances from the beam edge and with different field dimensions. One patient received radiotherapy with an inverted Y field and gonad doses were estimated using calibrated thermo-luminescent detector (TLD chips. Measured doses with the ion chamber were 7.1 and 3.5% of the mid-plane doses without a shield at 3 and 7.5 cm off-field respectively. These values decreased to 4.6 and 1.7% with the bottom shield alone, and to 1.7 and 0.8% with both bottom and top shields covering the ion chamber. The measured doses at the gonads during the patient′s treatment were 0.5-0.92% for the AP field (0.74 ± 0.17%, n = 5 and 0.5-1.2% for the PA field (0.88 ± 0.24%, n = 5. The dose received by the testis for the full course of treatment was 32 cGy (0.8% for a total mid-plane dose of 40 Gy. The first-scatter energy estimated at the gonads is around 1.14 MeV for a primary beam of 15 MV for a long axis dimension of 37 cm of primary field. During the patient′s treatment, the estimated absorbed doses at the gonads were comparable with reported values in similar treatments. The testicular shield reported in this study is of light weight and could be used conveniently in treatments of abdominal fields.

Fail-safe ion chamber errant beam detector tailored for personnel protection

International Nuclear Information System (INIS)

Plum, M.A.; Browman, A.A.; Brown, D.; Lee, D.M.; McCabe, C.W.

1989-01-01

This fail-safe ion chamber system is designed to be part of the personnel safety system (PSS) for the Los Alamos neutron Scattering Center (LANSCE) at the Los Alamos National Laboratory. Its job is to protect the occupants of the experimental areas from large radiation doses caused by errant beam conditions during beam transport from the Proton Storage Ring (PSR) to the LANSCE neutron spallation target. Due to limited shielding between the beam transport line and the experimental area only if the beam losses in the transport line are very low. The worst case beam spill scenario is calculated to result in a personnel exposure of about 0.01 Gys/s (1 rad/s). Although the preferred solution is to increase the bulk shielding between the beam line and the experimental area, the physical dimensions of the site do not permit an adequate amount of shielding to be added. The solution adopted is a layered system of three types of highly reliable detector systems: a current limiter system located in the beam line, a neutron detector system located in the experimental areas, and an ion chamber system located on the walls of the beam line tunnels. The ion chamber system is capable of shutting off the beam in less than 0.5 s, resulting in a worst case personnel exposure of 0.005 Gys (0.5 rad). 4 figs

The determination of self-powered neutron detector sensitivity on thermal and epithermal neutron flux densities

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

The coefficients of thermal and epithermal neutron flux density depression and self-shielding for the SPN detectors with vanadium, rhodium, silver and cobalt emitters are presented, (for cobalt SPN detectors the functions describing the absorbtion of neutrons along the emitter cross-section are also shown). Using these coefficients and previously published beta particle escape efficiencies, sensitivities are determined for the principal types of detectors produced by Les Cables de Lyon and SODERN companies. The experiments and their results verifying the validity of the theoretical work are described. (author)

Experimental apparatus with {sup 3}He detectors for photoneutron measurements greater than 8 MV; Aparato experimental de detector {sup 3}He, para medida fotonêutrons em acelerador com energia acima 8 MV

Energy Technology Data Exchange (ETDEWEB)

Silva, Marcelo Vilela da; Fontes, Gladson Silva; Cardoso, Domingos D’Oliveira, E-mail: marcelovilelasilva@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

2017-07-01

Particle technology has a profound impact on society. Its applications are well established mainly in the treatment of cancer and other diseases. The objective of this work was to develop an experimental apparatus with {sup 3}He detectors for photoneutron measurements greater than 8 MV. The apparatus allowed to obtain a multi energetic neutron measurements with detector and changing the different thicknesses of shields and for different dose rates. Additionally, an analytical model was developed, based on the approximation of the diffusion, for two groups of energy. The results indicated an increasing and linear behavior in the detector response. (author)

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.

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...

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

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.)

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

International Nuclear Information System (INIS)

McDonald, R.J.; Smith, A.R.; Norman, E.B.; Cowles, D.

1995-10-01

Following many years of productive work, the SuperHILAC and Bevalac accelerators at Lawrence Berkeley National Laboratory were closed, leaving thousands of concrete shielding blocks available for reuse or disposal. The process history of these blocks as shielding precludes free release pending radiological characterization. This paper presents a method for the rapid characterization of gamma-ray-emitting radioisotopes in large samples of earth-like materials: concrete shielding blocks in this case. Active regions are identified with a sensitive radiation-survey instrument and then examined in detail with a high-efficiency lead-shielded Ge spectrometer. Naturally-occurring gamma-ray emissions from the decays of uranium, thorium, and potassium are used to calibrate the spectrometer. A simple relationship exists between the observed counting rate in a characteristic gamma ray and the activity in the block. This method, taking only tens of minutes per sample at the nano-Curie/gram sensitivity level, replaces much of the expensive coring and laboratory analysis methods needed otherwise

X-ray shielding behaviour of kaolin derived mullite-barites ceramic

Science.gov (United States)

Ripin, A.; Mohamed, F.; Choo, T. F.; Yusof, M. R.; Hashim, S.; Ghoshal, S. K.

2018-03-01

Mullite-barite ceramic (MBC) is an emergent material for effective shielding of redundant ionizing radiation exposure. The composition dependent mechanical, thermal, and microstructure properties of MBC that makes MBC a high performing novel radiation shielding candidate remained unexplored. This paper examines the possibility of exploiting Malaysian kaolin (AKIM-35) and barite (BaSO4) derived ceramic (MBC) system for X-ray shielding operation. Using conventional pressing and sintering method six ceramic samples are prepared by mixing AKIM-35 with barite at varying contents (0, 10, 20, 30, 40 and 50 wt%). Synthesized pressed mixtures are calcined at 400 °C for 30 min and then sintered to 1300 °C for 120 min at a heating rate of 10 °C/min. Sintered samples are characterized via X-ray Diffraction (XRD), Field Emission Scanning Electron Microscope (FESEM), lead equivalent (LE), uniformity and dose reduction analyses. XRD pattern of prepared ceramics revealed the presence of monoclinic barium alumino-silicate (BAS) and orthorhombic mullite as major shielding phases together with other minor phase of barite and hexagonal quartz (SiO2) structures. Furthermore, FESEM images of ceramics (between 0 and 30 wt%) displayed the existence of compacted monoclinic plate of BAS and acicular mullite morphology (ceramics at 40 and 50 wt%). Radiation tests displayed the capacity of ceramics (at 0 and 10 wt%) to shield the X-ray radiation emanated at tube potential range of 50-120 kV. The highest radiation attenuation is ascertained at 70 kV where the dose is reduced remarkably between 99.11% and 97.42%. Ceramics at 0 and 10 wt% demonstrated the highest lead (Pb) equivalent thickness (LE) of 0.44 mm and 0.34 mm, respectively. It is established that such MBC may contribute towards the development of shielding material against ionizing radiation in diagnostic radiology (X-ray) dose range.

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

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.

Analysis of self-powered gamma ray detector with directional discrimination

International Nuclear Information System (INIS)

Levert, F.E.; Beyerlein, R.A.; Cox, S.A.

1979-01-01

The results of a combined Monte Carlo simulation and experimental investigation of the directional and energy dependent response of a self-powered gamma detector with a flat plate Pb-C central electrode are presented. The electron yield of the central electrode in a three dimensional mockup of the detector was calculated for photons of several discrete energies, emanating from an infinitely thin planar source, incident on the outer surface of the detector. Separate computations were done with the source facing the lead side and carbon side of the central electrode. Experimental measurements with a detector that closely matched the design used in the simulation were conducted in a graphite column next to a neutron leakage face of a low flux reactor. A localized gamma ray source was created by positioning a 235 U strip between the leakage face of the reactor and the detector. A comparison of results obtained in both cases showed good agreement. Also experimental measurements to determine the effect of the thickness of lead shielding surrounding the outer wall of the detector and space charge in the vacuum insulator between the central electrode and the inner wall on the response of the detector were performed. (Auth.)

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

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.

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)

Prospects for very large, sensitive water Cherenkov detectors for proton decay and neutrino oscillations search

International Nuclear Information System (INIS)

Cline, D.B.

1982-01-01

We discuss the possibility of constructing large water Cherenkov detectors with mass 10 5 to 10 6 tons that would be sensitive to a few hundred MeV - few GeV energy release. The 10 5 ton detector would be suitable for a search for certain proton decay modes whereas the 10 6 ton detector would act as an active shield for the proton decay detector and as a nu/sub e/, nu/sub μ/ and possibly nu/sub tau/ interaction detector. The neutrino physics would include a sensitive search for neutrino oscillations using atmospheric neutrinos. The location of this detector could be in the deep ocean near Hawaii or in a deep trench between Cuba and Haiti or perhaps deep lakes like Superior or Baikal if flexible containers are used

Note: Microelectrode-shielding tip for scanning probe electron energy spectroscopy

Science.gov (United States)

Huang, Wei; Li, Zhean; Xu, Chunkai; Liu, Jian; Xu, Chunye; Chen, Xiangjun

2018-04-01

We report a novel microelectrode-shielding tip (ME tip) for scanning probe electron energy spectroscopy (SPEES). The shielding effect of this tip is studied through comparing the detection efficiency with the normal tip by both experiment and simulation. The results show that the backscattering count rate detected by the SPEES instrument using the normal tip begins to decrease as the tip approaches to the sample surface within 21 μm, while that using the ME tip only starts to drop off within 1 μm. This indicates that the electron energy spectra can be measured with the ME tip at a much closer tip-sample distance. Furthermore, it is also demonstrated that the ME tip can be used to obtain topography of the sample surface in situ simultaneously.

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)

Development of a Cryogenic Radiation Detector for Mapping Radio Frequency Superconducting Cavity Field Emissions

CERN Document Server

Dotson, Danny W

2005-01-01

There is a relationship between field emissions in a Super Conducting RF cavity and the production of radiation (mostly X-rays). External (room temperature) detectors are shielded from the onset of low energy X-rays by the vacuum and cryogenic stainless steel module walls. An internal measuring system for mapping field emissions would assist scientists and engineers in perfecting surface deposition and acid washing module surfaces. Two measurement systems are undergoing cryogenic testing at JLab. One is an active CsI photodiode array and the second is an X-ray film camera. The CsI array has operated sucessfully in a cavity in liquid Helium but saturated at higher power due to scattering in the cavity. A shield with an aperature similar to the X-ray film detector is being designed for the next series of tests which will be completed before PAC-05.

Feasibility studies on large sample neutron activation analysis using a low power research reactor

International Nuclear Information System (INIS)

Gyampo, O.

2008-06-01

Instrumental neutron activation analysis (INAA) using Ghana Research Reactor-1 (GHARR-1) can be directly applied to samples with masses in grams. Samples weights were in the range of 0.5g to 5g. Therefore, the representativity of the sample is improved as well as sensitivity. Irradiation of samples was done using a low power research reactor. The correction for the neutron self-shielding within the sample is determined from measurement of the neutron flux depression just outside the sample. Correction for gamma ray self-attenuation in the sample was performed via linear attenuation coefficients derived from transmission measurements. Quantitative and qualitative analysis of data were done using gamma ray spectrometry (HPGe detector). The results of this study on the possibilities of large sample NAA using a miniature neutron source reactor (MNSR) show clearly that the Ghana Research Reactor-1 (GHARR-1) at the National Nuclear Research Institute (NNRI) can be used for sample analyses up to 5 grams (5g) using the pneumatic transfer systems.

Ultra Low Level Environmental Neutron Measurements Using Superheated Droplet Detectors

Energy Technology Data Exchange (ETDEWEB)

Fernandes, A.C. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal); Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Felizardo, M.; Girard, T.A.; Kling, A.; Ramos, A.R. [Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Marques, J.G.; Prudencio, M.I.; Marques, R.; Carvalho, F.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal)

2015-07-01

Through the application of superheated droplet detectors (SDDs), the SIMPLE project for the direct search for dark matter (DM) reached the most restrictive limits on the spin-dependent sector to date. The experiment is based on the detection of recoils following WIMP-nuclei interaction, mimicking those from neutron scattering. The thermodynamic operation conditions yield the SDDs intrinsically insensitive to radiations with linear energy transfer below ∼150 keVμm{sup -1} such as photons, electrons, muons and neutrons with energies below ∼40 keV. Underground facilities are increasingly employed for measurements in a low-level radiation background (DM search, gamma-spectroscopy, intrinsic soft-error rate measurements, etc.), where the rock overburden shields against cosmic radiation. In this environment the SDDs are sensitive only to α-particles and neutrons naturally emitted from the surrounding materials. Recently developed signal analysis techniques allow discrimination between neutron and α-induced signals. SDDs are therefore a promising instrument for low-level neutron and α measurements, namely environmental neutron measurements and α-contamination assays. In this work neutron measurements performed in the challenging conditions of the latest SIMPLE experiment (1500 mwe depth with 50-75 cm water shield) are reported. The results are compared with those obtained by detailed Monte Carlo simulations of the neutron background induced by {sup 238}U and {sup 232}Th traces in the facility, shielding and detector materials. Calculations of the neutron energy distribution yield the following neutron fluence rates (in 10{sup -8} cm{sup -2}s{sup -1}): thermal (<0.5 eV): 2.5; epithermal (0.5 eV-100 keV): 2.2; fast (>1 MeV): 3.9. Signal rates were derived using standard cross sections and codes routinely employed in reactor dosimetry. The measured and calculated neutron count rates per unit of active mass were 0.15 ct/kgd and 0.33 ct/kg-d respectively. As the major

An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

Science.gov (United States)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

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.

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)

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.

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

Shielding structure analysis for LSDS facility

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Shielding structure analysis for LSDS facility

International Nuclear Information System (INIS)

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

2014-01-01

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

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.)

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

Science.gov (United States)

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

2018-06-01

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

Current approaches to geochemical reconnaissance for uranium in the Canadian Shield

International Nuclear Information System (INIS)

Cameron, E.M.; Hornbrook, E.H.W.

1976-01-01

Wide-interval geochemical reconnaissance is currently being carried out over large areas of the Canadian Shield by the Geological Survey of Canada. This work is in support of the federal-provincial Uranium Reconnaissance Program. The paper reviews the methodology employed for this reconnaissance to outline areas of enhanced potential for uranium and other mineral commodities. The generally low relief of the Shield virtually restricts wide-interval reconnaissance to the mobile elements that can travel some distance in solution. The high mobility of uranium, particularly in waters of neutral pH, makes it one of the most suitable types of mineralization for detection. For much of the Shield the most appropriate sampling media are lake sediments and lake waters. Centre-lake sediments are homogeneous, and are readily and economically collected utilizing helicopter support. They provide a good response for uranium and a variety of other indicator elements. Waters collected at the same sample sites are a useful supplement in the search for uranium mineralization, particularly in carbonate terrain. However, because of the very low content of uranium in many Shield waters, they present as yet unresolved problems of analysis and storage. The influence of organic material, iron and manganese on the uranium content of lake sediments are examined. Their scavenging influence has been found to be significant only at their lower concentration levels, particularly for organic matter and iron. For the majority of centre-lake samples the effect is negligible. The choice of sample interval is related to total survey costs, to speed of coverage, and to ability to detect uranium and other types of mineralization. For current G.S.C. programs a sample density of one per five square mile appears optimal. The results are presented as l:250,000 symbol maps that are prepared largely by computer, and as 1:1,000,000 contoured compilation maps. (author)

HZE dosimetry in space using plastic track detectors

CERN Document Server

Kopp, J; Reitz, G; Enge, W

1999-01-01

Plastic nuclear track detectors were used to measure the contribution of High charge Z and energy E (HZE) particles to the radiation exposure of manned space missions. Results from numerous space missions in the orbit planned for the International Space Station are compared. The measurements cover the declining phase of the last solar cycle during the past 7 years and various shielding conditions inside the US Space Shuttle and the Russian MIR-station.

Control of the neutron detector count rate by optical imaging

International Nuclear Information System (INIS)

Roquemore, A.L.; Johnson, L.C.

1992-01-01

The signal processing electronics used for the NE451 detectors on the TFTR multichannel neutron collimator are presently showing saturation effects at high counting rates equivalent to neutron yields of ∼10 16 n/s. While nonlinearity due to pulse pileup can be corrected for in most present TFTR experiments, additional steps are required for neutron source strengths above ∼3x10 16 n/s. These pulse pileup effects could be reduced by inserting sleeves in the collimator shielding to reduce the neutron flux in the vicinity of the detectors or by reducing the volume of detector exposed to the flux. We describe a novel method of avoiding saturation by optically controlling the number neutron events processed by the detector electronics. Because of the optical opacity of the ZnS-plastic detectors such as NE451, photons from a proton-recoil scintillation arise from a spatially localized area of the detector. By imaging a selected portion of the detector onto a photomultiplier, we reduce the effective volume of the detector in a controllable, reversible way. A prototype system, consisting of a focusing lens, a field lens, and a variable aperture, has been constructed. Results of laboratory feasibility tests are presented

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''

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)

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

LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

International Nuclear Information System (INIS)

Kyllingstad, G.

1977-01-01

The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

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

The detector calibration system for the CUORE cryogenic bolometer array

Energy Technology Data Exchange (ETDEWEB)

Cushman, Jeremy S., E-mail: jeremy.cushman@yale.edu [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Dally, Adam [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Davis, Christopher J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Ejzak, Larissa; Lenz, Daniel [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Lim, Kyungeun E. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Heeger, Karsten M., E-mail: karsten.heeger@yale.edu [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Maruyama, Reina H. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Nucciotti, Angelo [Dipartimento di Fisica, Università di Milano-Bicocca, Milano I-20126 (Italy); INFN – Sezione di Milano Bicocca, Milano I-20126 (Italy); Sangiorgio, Samuele [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Wise, Thomas [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Department of Physics, University of Wisconsin, Madison, WI 53706 (United States)

2017-02-01

The Cryogenic Underground Observatory for Rare Events (CUORE) is a ton-scale cryogenic experiment designed to search for neutrinoless double-beta decay of {sup 130}Te and other rare events. The CUORE detector consists of 988 TeO{sub 2} bolometers operated underground at 10 mK in a dilution refrigerator at the Laboratori Nazionali del Gran Sasso. Candidate events are identified through a precise measurement of their energy. The absolute energy response of the detectors is established by the regular calibration of each individual bolometer using gamma sources. The close-packed configuration of the CUORE bolometer array combined with the extensive shielding surrounding the detectors requires the placement of calibration sources within the array itself. The CUORE Detector Calibration System is designed to insert radioactive sources into and remove them from the cryostat while respecting the stringent heat load, radiopurity, and operational requirements of the experiment. This paper describes the design, commissioning, and performance of this novel source calibration deployment system for ultra-low-temperature environments.

Boosted Dark Matter Quarrying at Surface Neutrino Detectors arXiv

CERN Document Server

Kim, Doojin; Park, Jong-Chul; Shin, Seodong

We propose the idea of "Earth Shielding" to reject cosmic-ray backgrounds, in the search for boosted dark matter at surface neutrino detectors, resulting in the enhancement of the signal-to-background ratio. The identification of cosmic-originating rare signals, especially lacking features, at surface detectors is often considered hopeless due to a vast amount of cosmic-ray-induced background, hence underground experiments are better motivated to avoid such a challenge. We claim that surface detectors can attain remarkable sensitivities to even featureless signals, once restricting to events coming through the Earth from the opposite side of the detector location for the signals leaving appreciable tracks from which the source direction is inferred. By doing so, potential backgrounds in the signal region of interest can be substantially suppressed. To validate our claim, we study experimental reaches at several surface experiments such as SBN Program (MicroBooNE, ICARUS, and SBND) and ProtoDUNE for elastic bo...

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.

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.)

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)

Remote quantitative analysis of cerium through a shielding window by stand-off laser-induced breakdown spectroscopy

International Nuclear Information System (INIS)

Gong, Yongdeuk; Choi, Daewoong; Han, Bo-Young; Yoo, Jonghyun; Han, Song-Hee; Lee, Yonghoon

2014-01-01

Highlights: • Remote LIBS analysis of cerium in the samples located behind a shielding window. • Effects of a shielding window on the remote LIBS analysis were investigated. • Multivariate analysis improves the calibration quality. - Abstract: Laser-Induced Breakdown Spectroscopy (LIBS) has been considered in many applications in nuclear industry. LIBS can be an ideal technique for analyzing the inaccessible nuclear materials typically located behind a shielding window. We report the effect of optical transmittance of the shielding window on the analytical performances of stand-off LIBS for the preliminary surrogate sample of demonstration pyrochemical process, a mixture of cerium oxide (CeO 2 ) and potassium chloride (KCl). A pulsed laser beam was focused on the surface of the sample located 1.45 m away from the stand-off LIBS device. The laser-induced plasma emission was collected through a Schmidt–Cassegrain telescope. LIBS spectra were obtained in an open path and through the shielding window. Univariate calibration curves were obtained using the integrated area of partially resolved Ce I and II lines. The limits of detection (LOD) for Ce were estimated to be 0.046 and 0.061 wt.% for the open-path and through-window analysis, respectively. We found that the through-window LOD is mainly influenced by the optical transmittance of the shielding window and therefore, the through-window LOD can be predicted from the open-path LOD and the optical transmittance of the shielding window. Also, multivariate calibration using partial least squares regression was successfully applied. The quality of calibration could be improved by the multivariate analysis

Remote quantitative analysis of cerium through a shielding window by stand-off laser-induced breakdown spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Gong, Yongdeuk [Department of Chemistry, Mokpo National University, Jeonnam 534-729 (Korea, Republic of); Choi, Daewoong [Department of Chemistry, Mokpo National University, Jeonnam 534-729 (Korea, Republic of); Korea Atomic Energy Research Institute, P.O. Box 255, Yuseong, Daejeon 305-353 (Korea, Republic of); Han, Bo-Young, E-mail: byhan@kaeri.re.kr [Korea Atomic Energy Research Institute, P.O. Box 255, Yuseong, Daejeon 305-353 (Korea, Republic of); Yoo, Jonghyun [Applied Spectra, 46665 Fremont Boulevard, Fremont, CA 94538 (United States); Han, Song-Hee [Division of Maritime Transportation System, Mokpo National Maritime University, Jeonnam 530-729 (Korea, Republic of); Lee, Yonghoon, E-mail: yhlee@mokpo.ac.kr [Department of Chemistry, Mokpo National University, Jeonnam 534-729 (Korea, Republic of)

2014-10-15

Highlights: • Remote LIBS analysis of cerium in the samples located behind a shielding window. • Effects of a shielding window on the remote LIBS analysis were investigated. • Multivariate analysis improves the calibration quality. - Abstract: Laser-Induced Breakdown Spectroscopy (LIBS) has been considered in many applications in nuclear industry. LIBS can be an ideal technique for analyzing the inaccessible nuclear materials typically located behind a shielding window. We report the effect of optical transmittance of the shielding window on the analytical performances of stand-off LIBS for the preliminary surrogate sample of demonstration pyrochemical process, a mixture of cerium oxide (CeO{sub 2}) and potassium chloride (KCl). A pulsed laser beam was focused on the surface of the sample located 1.45 m away from the stand-off LIBS device. The laser-induced plasma emission was collected through a Schmidt–Cassegrain telescope. LIBS spectra were obtained in an open path and through the shielding window. Univariate calibration curves were obtained using the integrated area of partially resolved Ce I and II lines. The limits of detection (LOD) for Ce were estimated to be 0.046 and 0.061 wt.% for the open-path and through-window analysis, respectively. We found that the through-window LOD is mainly influenced by the optical transmittance of the shielding window and therefore, the through-window LOD can be predicted from the open-path LOD and the optical transmittance of the shielding window. Also, multivariate calibration using partial least squares regression was successfully applied. The quality of calibration could be improved by the multivariate analysis.

Summary report of neutral detector subgroup

International Nuclear Information System (INIS)

Akerlof, C.; Bensinger, J.; Donaldson, G.

1977-01-01

The advantages and disadvantages of calorimeters for use at ISABELLE are discussed including hadron calorimeters, liquid argon calorimeters, uranium scintillator calorimeters, electromagnetic shower detectors, and iron plate calorimeters. A calorimeter is described that is suitable for electromagnetic and hadronic events and is constructed with a front section of narrow Pb strips to start and develop the electromagnetic shower. A section of U towers to contain hadronic showers follows. This calorimeter would shield out U radioactivity and measure the energy and position of all particles except neutrinos and muons

Investigation of the use of Galena concrete in electromagnetic radiation shielding

International Nuclear Information System (INIS)

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

2011-01-01

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

Muon flux measurement with silicon detectors in the CERN neutrino beams

International Nuclear Information System (INIS)

Heijne, H.M.

1983-01-01

The present work mainly describes the 'Neutrino Flux Monitoring' system (NFM), which has been built for the 400-GeV Super Proton Synchrotron (SPS) neutrino beams. A treatment is given of some general subjects related to the utilization of silicon detectors and the properties of high-energy muons. Energy loss of minimal-ionizing particles, which has to be distinguished from energy deposition in the detector, is considered. Secondary radiation, also called 'spray', consisting of 'delta rays' and other cascade products, is shown to play an important role in the muon flux measurement inside a shield, especially for muons of high energy (> 100 GeV). Radiation induced damage in the detectors, which determines the long term performance, is discussed. The relation between the detector response and the real muon flux is determined. The use of NFM system for on-line beam monitoring is described. (Auth.)

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

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

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.)

Magnetic shield effect simulation of superconducting film shield covering directly coupled HTS dc-SQUID magnetometer

International Nuclear Information System (INIS)

Terauchi, N.; Noguchi, S.; Igarashi, H.

2011-01-01

A superconducting film shield over a SQUID ring improves the robustness of the SQUID with respect to magnetic noise. Supercurrent in the SQUID magnetometer and the superconducting film shield were simulated. The superconducting film shield reduces the influence of the external magnetic field on the SQUID ring. An HTS SQUID is a high sensitive magnetic sensor. In recent years, the HTS SQUID is widely used in various applications. In some applications, high robustness with respect to magnetic noise is required to realize stable operation at outside of a magnetic shielding room. The target of this paper is a directly coupled HTS dc-SQUID magnetometer. To enhance the robustness of the SQUID magnetometer, use of a superconducting thin film shield has been proposed. The magnetic field directly penetrating the SQUID ring causes the change of the critical current of Josephson junction, and then the SQUID magnetometer transitions into inoperative state. In order to confirm the magnetic shield effect of the superconducting film shield, electromagnetic field simulation with 3D edge finite element method was performed. To simulate the high temperature superconductor, E-J characteristics and c-axis anisotropy are considered. To evaluate the effect of the superconducting film shield, an external magnetic field which is supposed to be a magnetic noise is applied. From the simulation results, the time transition of the magnetic flux penetrating the SQUID ring is investigated and the effect of the superconducting film shield is confirmed. The amplitude of the magnetic flux penetrating the SQUID ring can be reduced to about one-sixth since the superconducting film shield prevents the magnetic noise from directly penetrating the SQUID ring.

Passive magnetic shielding in MRI-Linac systems

Science.gov (United States)

Whelan, Brendan; Kolling, Stefan; Oborn, Brad M.; Keall, Paul

2018-04-01

Passive magnetic shielding refers to the use of ferromagnetic materials to redirect magnetic field lines away from vulnerable regions. An application of particular interest to the medical physics community is shielding in MRI systems, especially integrated MRI-linear accelerator (MRI-Linac) systems. In these systems, the goal is not only to minimize the magnetic field in some volume, but also to minimize the impact of the shield on the magnetic fields within the imaging volume of the MRI scanner. In this work, finite element modelling was used to assess the shielding of a side coupled 6 MV linac and resultant heterogeneity induced within the 30 cm diameter of spherical volume (DSV) of a novel 1 Tesla split bore MRI magnet. A number of different shield parameters were investigated; distance between shield and magnet, shield shape, shield thickness, shield length, openings in the shield, number of concentric layers, spacing between each layer, and shield material. Both the in-line and perpendicular MRI-Linac configurations were studied. By modifying the shield shape around the linac from the starting design of an open ended cylinder, the shielding effect was boosted by approximately 70% whilst the impact on the magnet was simultaneously reduced by approximately 10%. Openings in the shield for the RF port and beam exit were substantial sources of field leakage; however it was demonstrated that shielding could be added around these openings to compensate for this leakage. Layering multiple concentric shield shells was highly effective in the perpendicular configuration, but less so for the in-line configuration. Cautious use of high permeability materials such as Mu-metal can greatly increase the shielding performance in some scenarios. In the perpendicular configuration, magnetic shielding was more effective and the impact on the magnet lower compared with the in-line configuration.

The effect of cadmium shielding on the spatial neutron flux distribution inside one of the outer irradiation sites

International Nuclear Information System (INIS)

Shaaban, I.

2009-06-01

A permanent epithermal neutron irradiation facility was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using the cadmium (cylindrical vial 1.0 mm in thickness, 38.50 mm in diameter and 180 mm in length) as thermal neutron shielding material, for a permanent epithermal neutron activation analysis (ENAA). This site was designed by shielding the internal surface of the aluminum tube of the first outer irradiation site in the MNSR reactor. I was used the activation detectors 0.1143% Au-Al alloy foils with 0.1 mm thickness and 2.0 mm diameter for measurement the thermal neutron flux, epithermal and R c d=A b are/A c over ratio in the outer irradiation site. Distribution of the thermal neutron flux in the outer irradiation capsule has been found numerically using MCNP-4C code with and without cadmium shield, and experimentally by irradiating five copper wires using the outer irradiation capsule. Good agreements were obtained between the calculated and the measured results. (author)

New shielding configurations for a simultaneous PET/MRI scanner at 7T

Science.gov (United States)

Peng, Bo J.; Wu, Yibao; Cherry, Simon R.; Walton, Jeffrey H.

2014-02-01

Understanding sources of electromagnetic interference are important in designing any electronic system. This is especially true when combining positron emission tomography (PET) and magnetic resonance imaging (MRI) in a multimodality system as coupling between the subsystems can degrade the performance of either modality. For this reason, eliminating radio frequency (RF) interference and gradient-induced eddy currents have been major challenges in building simultaneous hybrid PET/MRI systems. MRI requires negligible RF interference at the Larmor resonance frequency, while RF interference at almost any frequency may corrupt PET data. Moreover, any scheme that minimizes these interactions would, ideally, not compromise the performance of either subsystem. This paper lays out a plan to resolve these problems. A carbon fiber composite material is found to be a good RF shield at the Larmor frequency (300 MHz in this work) while introducing negligible gradient eddy currents. This carbon fiber composite also provides excellent structural support for the PET detector components. Low frequency electromagnetic radiation (81 kHz here) from the switching power supplies of the gradient amplifiers was also found to interfere with the PET detector. Placing the PET detector module between two carbon fiber tubes and grounding the inner carbon fiber tube to the PET detector module ground reduced this interference. Further reductions were achieved by adding thin copper (Cu) foil on the outer carbon fiber case and electrically grounding the PET detector module so that all 3 components had a common ground, i.e. with the PET detector in an electrostatic cage. Finally, gradient switching typical in MRI sequences can result in count losses in the particular PET detector design studied. Moreover, the magnitude of this effect depends on the location of the detector within the magnet bore and which MRI gradient is being switched. These findings have a bearing on future designs of PET

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

ASTRO-H CdTe detectors proton irradiation at PIF

International Nuclear Information System (INIS)

Limousin, O.; Renaud, D.; Horeau, B.; Dubos, S.; Laurent, P.; Lebrun, F.; Chipaux, R.; Boatella Polo, C.; Marcinkowski, R.; Kawaharada, M.; Watanabe, S.; Ohta, M.; Sato, G.; Takahashi, T.

2015-01-01

Asbstract: The French Atomic Energy Commission (CEA), with the support of the European Space Agency (ESA), is partner of the Soft Gamma-Ray Detector (SGD) and the Hard X-ray Imager (HXI) onboard the 6th Japanese X-ray scientific satellite ASTRO-H (JAXA) initiated by the Institute of Space and Astronautical Science (ISAS). Both scientific instruments, one hosting a series of Compton Gamma Cameras and the other being a focal plane of a grazing incidence mirror telescope in the hard X-ray domain, are equipped with Cadmium Telluride based detectors. ASTRO-H will be operated in a Low Earth Orbit with a 31° inclination at ~550 km altitude, thus passing daily through the South Atlantic Anomaly radiation belt, a specially harsh environment where the detectors are suffering the effect of the interaction with trapped high energy protons. As CdTe detector performance might be affected by the irradiation, we investigate the effect of the accumulated proton fluence on their spectral response. To do so, we have characterized and irradiated representative samples of SGD and HXI detector under different conditions. The detectors in question, from ACRORAD, are single-pixels having a size of 2 mm by 2 mm and 750 µm thick. The Schottky contact is either made of an Indium or Aluminum for SGD and HXI respectively. We ran the irradiation test campaign at the Proton Irradiation Facility (PIF) at PSI, and ESA approved equipment to evaluate the radiation hardness of flight hardware. We simulated the proton flux expected on the sensors over the entire mission, and secondary neutrons flux due to primary proton interactions into the surrounding BGO active shielding. We eventually characterized the detector response evolution, emphasizing each detector spectral response as well as its stability by studying the so-called Polarization effect. The latter is provoking a spectral response degradation against time as a charge accumulation process occurs in Schottky type CdTe sensors. In this paper

Characteristics of the quarry as shielding for "2"4"1AmBe neutrons and monoenergetic photons

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A.; Rodriguez R, J. A.; Juarez A, C. A.

2016-09-01

Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of "2"4"1AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10"-"3 to 10"-"5 MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the "2"4"1AmBe neutron spectrum as a function of quarry thickness, as well as the E_R_O_T and H(10) transmission curves. (Author)

EBT-P gamma-ray-shielding analysis

International Nuclear Information System (INIS)

Gohar, Y.

1983-01-01

First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

Radiation Build-Up Of High Energy Gamma In Shielding Of High Atomic Number

International Nuclear Information System (INIS)

Yuliati, Helfi; Akhadi, Mukhlis

2000-01-01

Research to observe effect of radiation build-up factor (b) in iron (Fe) and lead (Pb) for high energy gamma shielding from exp.137 Cs (E gamma : 662 keV) and exp.60 Co (E gamma : 1332 keV) sources has been carried out. Research was conducted bt counting of radiation intensity behind shielding with its thickness vary from 1 to 5 times of half value thickness (HVT). NaI (TI) detector which connected to multi channel analyzer (MCA) was used for the counting. Calculation result show that all of b value are near to 1 (b∼1) both for Fe and Pb. Without inserting b in calculation, from the experiment it was obtained HVT value of Fe for high gamma radiation of 662 and 1332 keV were : (12,94 n 0,03) mm and (17,33 n 0,01) mm with their deviation standards were 0,2% and 0,06% respectively. Value of HVT for Pb with the same energy were : (6,31 n 0,03) mm and (11,86 n 0,03) mm with their deviation standars were : 0,48% and 0,25% respectively. HVL concept could be applied directly to estimate shielding thickness of high atomic number of high energy gamma radiation, without inserting correction of radiation build-up factor

A comparison of shielding calculation methods for multi-slice computed tomography (CT) systems

International Nuclear Information System (INIS)

Cole, J A; Platten, D J

2008-01-01

Currently in the UK, shielding calculations for computed tomography (CT) systems are based on the BIR-IPEM (British Institute of Radiology and Institute of Physics in Engineering in Medicine) working group publication from 2000. Concerns have been raised internationally regarding the accuracy of the dose plots on which this method depends and the effect that new scanner technologies may have. Additionally, more recent shielding methods have been proposed by the NCRP (National Council on Radiation Protection) from the USA. Thermoluminescent detectors (TLDs) were placed in three CT scanner rooms at different positions for several weeks before being processed. Patient workload and dose data (DLP: the dose length product and mAs: the tube current-time product) were collected for this period. Individual dose data were available for more than 95% of patients scanned and the remainder were estimated. The patient workload data were used to calculate expected scatter radiation for each TLD location by both the NCRP and BIR-IPEM methods. The results were then compared to the measured scattered radiation. Calculated scattered air kerma and the minimum required lead shielding were found to be frequently overestimated compared to the measured air kerma, on average almost five times the measured scattered air kerma.

Simulation of a method for determining one-dimensional {sup 137}Cs distribution using multiple gamma spectroscopic measurements with an adjustable cylindrical collimator and center shield

Energy Technology Data Exchange (ETDEWEB)

Whetstone, Z.D.; Dewey, S.C. [Radiological Health Engineering Laboratory, Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, 1906 Cooley Building, Ann Arbor, MI 48109-2104 (United States); Kearfott, K.J., E-mail: kearfott@umich.ed [Radiological Health Engineering Laboratory, Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, 1906 Cooley Building, Ann Arbor, MI 48109-2104 (United States)

2011-05-15

With multiple in situ gamma spectroscopic measurements obtained with an adjustable cylindrical collimator and a circular shield, the arbitrary one-dimensional distribution of radioactive material can be determined. The detector responses are theoretically calculated, field measurements obtained, and a system of equations relating detector response to measurement geometry and activity distribution solved to estimate the distribution. This paper demonstrates the method by simulating multiple scenarios and providing analysis of the system conditioning.

Cryogenic readout techniques for germanium detectors

Energy Technology Data Exchange (ETDEWEB)

Benato, G. [University of Zurich, (Switzerland); Cattadori, C. [INFN - Milano Bicocca, (Italy); Di Vacri, A. [INFN LNGS, (Italy); Ferri, E. [Universita Milano Bicocca/INFN Milano Bicocca, (Italy); D' Andrea, V.; Macolino, C. [GSSI/INFN LNGS, (Italy); Riboldi, S. [Universita degli Studi di Milano/INFN Milano, (Italy); Salamida, F. [Universita Milano Bicocca/INFN Milano Bicocca, (Italy)

2015-07-01

High Purity Germanium detectors are used in many applications, from nuclear and astro-particle physics, to homeland security or environment protection. Although quite standard configurations are often used, with cryostats, charge sensitive amplifiers and analog or digital acquisition systems all commercially available, it might be the case that a few specific applications, e.g. satellites, portable devices, cryogenic physics experiments, etc. also require the development of a few additional or complementary techniques. An interesting case is for sure GERDA, the Germanium Detector Array experiment, searching for neutrino-less double beta decay of {sup 76}Ge at the Gran Sasso National Laboratory of INFN - Italy. In GERDA the entire detector array, composed of semi-coaxial and BEGe naked crystals, is operated suspended inside a cryostat filled with liquid argon, that acts not only as cooling medium and but also as an active shield, thanks to its scintillation properties. These peculiar circumstances, together with the additional requirement of a very low radioactive background from all the materials adjacent to the detectors, clearly introduce significant constraints on the design of the Ge front-end readout electronics. All the Ge readout solutions developed within the framework of the GERDA collaboration, for both Phase I and Phase II, will be briefly reviewed, with their relative strength and weakness compared together and with respect to ideal Ge readout. Finally, the digital processing techniques developed by the GERDA collaboration for energy estimation of Ge detector signals will be recalled. (authors)

SUPERCOLLIDER: A GEM of a detector

International Nuclear Information System (INIS)

Anon.

1993-01-01

Now being prepared as a major experimental facility for the 87- kilometre Superconducting Supercollider (SSC) being built in Ellis County, Texas, is the GEM detector project. GEM thus becomes the companion to the Solenoidal Detector Collaboration (SDC), the first major SSC detector to emerge (March 1992, page 13). This is in keeping with the SSC Laboratory's aim of two major detectors with overlapping and complementary strengths. GEM is designed to observe all SSC signatures, with emphasis on precise measurement of electrons, photons and muons. Hence the name GEM - ''Gammas, Electrons and Muons.'' Design goals are clean signatures for leptons, jets, and missing transverse energy, maximum sensitivity to narrow resonances, and low backgrounds. Also important is maintaining significant capability at high luminosity (10 34 cm -2 s -1 ). GEM has some distinctive features. A key concept is the exterior magnet, surrounding all detector elements. Inside the magnet are a muon tracking system, a precision calorimeter, and a compact central tracker. This allows the muon momentum to be measured the air of the radiation shielded area outside the thick calorimeter, giving both high precision and robustness at high luminosity. A large magnet gives a large lever arm (at least 4 m) for precise muon momentum measurement. Placing the magnet outside also minimizes the material between tracker and calorimeters, so that the calorimeters are limited only by their inherent resolutions

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)

A low cost network of spectrometer radiation detectors based on the ArduSiPM a compact transportable Software/Hardware Data Acquisition system with Arduino DUE

International Nuclear Information System (INIS)

Bocci, Valerio; Chiodi, Giacomo; Iacoangeli, Francesco; Nuccetelli, Massimo; Recchia, Luigi

2015-01-01

The necessity to use Photo Multipliers (PM) as light detector limited in the past the use of crystals in radiation handled device preferring the Geiger approach. The Silicon Photomultipliers (SiPMs) are very small and cheap, solid photon detectors with good dynamic range and single photon detection capability, they are usable to supersede cumbersome and difficult to use Photo Multipliers (PM). A SiPM can be coupled with a scintillator crystal to build efficient, small and solid radiation detector. A cost effective and easily replicable Hardware software module for SiPM detector readout is made using the ArduSiPM solution. The ArduSiPM is an easily battery operable handled device using an Arduino DUE (an open Software/Hardware board) as processor board and a piggy-back custom designed board (ArduSiPM Shield), the Shield contains all the blocks features to monitor, set and acquire the SiPM using internet network. (authors)

A low cost network of spectrometer radiation detectors based on the ArduSiPM a compact transportable Software/Hardware Data Acquisition system with Arduino DUE

Energy Technology Data Exchange (ETDEWEB)

Bocci, Valerio; Chiodi, Giacomo; Iacoangeli, Francesco; Nuccetelli, Massimo; Recchia, Luigi [INFN Sezione di Roma, P.le Aldo moro 2, Rome, I-00185 (Italy)

2015-07-01

The necessity to use Photo Multipliers (PM) as light detector limited in the past the use of crystals in radiation handled device preferring the Geiger approach. The Silicon Photomultipliers (SiPMs) are very small and cheap, solid photon detectors with good dynamic range and single photon detection capability, they are usable to supersede cumbersome and difficult to use Photo Multipliers (PM). A SiPM can be coupled with a scintillator crystal to build efficient, small and solid radiation detector. A cost effective and easily replicable Hardware software module for SiPM detector readout is made using the ArduSiPM solution. The ArduSiPM is an easily battery operable handled device using an Arduino DUE (an open Software/Hardware board) as processor board and a piggy-back custom designed board (ArduSiPM Shield), the Shield contains all the blocks features to monitor, set and acquire the SiPM using internet network. (authors)

LORINE: Neutron emission Locator by SOI detectors

Energy Technology Data Exchange (ETDEWEB)

Hamrita, H.; Kondrasovs, V.; Borbotte, J. M.; Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electronique, F-91191 Gif-sur-Yvette Cedex (France); Saurel, N. [CEA, DAM, VALDUC, F-21120 Is sur Tille (France)

2009-07-01

The aim of this work is to develop a fast Neutron Emission Locator based on silicon on Insulator detector (LORINE). This locator can be used in the presence of significant flux of gamma radiation. LORINE was developed to locate areas containing a significant amount of actinide during the dismantling operations of equipment. From the results obtained in laboratory, we have proposed the prototype of neutron emission locator as follows: the developed design consists of 5 SOI (Silicon-on-insulator) detectors (1*1 cm{sup 2}) with their charge preamplifiers and their respective converters. All are installed on 5 faces of a boron polyethylene cube (5*5*5 cm{sup 3}). This cube plays the role of neutron shielding between the several detectors. The design must be so compact for use in glove boxes. An electronic card based on micro-controller has been made to control sensors and to send the necessary information to the computer. Location of fast neutron sources does not yet exist in a so compact design and it can be operated in the presence of very important gamma radiation flux

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

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)

Calculation of spatial weight functions for WWER-440 ex-core neutron detectors

International Nuclear Information System (INIS)

Csom, Gy.; Czifrus, Sz.; Feher, S.; Berki, T.

2001-01-01

The objective of the work presented in this paper was determination of a spatial weight function for WWER-440 ex-core detectors to be used for the interpretation of reload startup rod drop measurements. In view of the complexity of the geometry of the core as well as the detector, furthermore the presence of a cavity between the vessel and the concrete shield, Monte Carlo calculations were applied. In spite of the fact that in the corresponding literature the use of adjoint methods dominates, in the present case the forward method was chosen and implemented using MCNP4C (Authors)

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

Monte Carlo validation and optimisation of detector packaging for spectroscopic dosimetry for in vivo urethral dosimetry during low dose rate brachytherapy

International Nuclear Information System (INIS)

Nourbehesht, L.K.; Cutajar, D.L.; Guatelli, S.; Rosenfeld, A.B.

2015-01-01

The urethral mini-dosimeter, developed by the Centre for Medical Radiation Physics, University of Wollongong, uses spectroscopic dosimetry to provide real time point dose measurements along the urethra during low dose rate prostate brachytherapy. Spectroscopic dosimetry uses the measured spectrum of the treatment isotope to estimate the dose rate at the point of measurement, however, the silicon mini-detectors employed in the urethral mini-dosimeter require water proof encapsulation which must be capable of providing electromagnetic shielding without greatly increasing the size of the probe. The introduction of non-tissue equivalent materials within the encapsulation can change the spectrum of radiation incident on the detector, which may influence the application of spectroscopic dosimetry within the urethral dosimeter. The Monte Carlo code Geant4 was adopted to study the effect of encapsulation on the operation of the urethral mini-dosimeter, as well as to determine whether an appropriate thickness of aluminium shielding was possible for electromagnetic screening. The depth dose response and angular dependence of the urethral mini-dosimeter with three thicknesses of aluminium shielding (20, 50, 100 µm) was compared with the urethral mini-dosimeter without aluminium shielding. The aluminium shielding had the effect of increasing the depth dose response (up to 3 % within 30 mm and up to 5 % within 50 mm), slightly reduced the azimuth angular dependence and slightly increased the polar angular dependence. The 100 µm thick shielding provided the least azimuth angular dependence (±2 %) and provided a polar angular dependence of ±1.4 % within the angles of −45° to 45°.

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