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Sample records for gamma neutron assay

  1. Quantitative radiological characterization of waste. Integration of gamma spectrometry and passive/active neutron assay

    Energy Technology Data Exchange (ETDEWEB)

    Simone, Gianluca; Mauro, Egidio; Gagliardi, Filippo; Gorello, Edoardo [Nucleco S.p.A., Rome (Italy)

    2016-06-15

    The radiological characterization of drums through Non-Destructive Assay (NDA) techniques commonly relies on gamma spectrometry. This paper introduces the procedure developed in Nucleco for the NDA radiological characterization of drums when the presence of Special Nuclear Material (SNM) is expected/observed. The procedure is based on the integration of a gamma spectrometry in SGS mode (Segmented Gamma Scanner) and a passive/active neutron assay. The application of this procedure is discussed on a real case of drums. The extension of the integration procedure to other gamma spectrometry systems is also discussed.

  2. Radioactive waste package assay facility. Volume 2. Investigation of active neutron and active gamma interrogation

    International Nuclear Information System (INIS)

    Bailey, M.; Bunce, L.J.; Findlay, D.J.S.; Jolly, J.E.; Parsons, T.V.; Sene, M.R.; Swinhoe, M.T.

    1992-01-01

    Volume 2 of this report describes the theoretical and experimental work carried out at Harwell on active neutron and active gamma interrogation of 500 litre cemented intermediate level waste drums. The design of a suitable neutron generating target in conjunction with a LINAC was established. Following theoretical predictions of likely neutron responses, an experimental assay assembly was built. Responses were measured for simulated drums of ILW, based on CAGR, Magnox and PCM wastes. Good correlations were established between quantities of 235 -U, nat -U and D 2 O contained in the drums, and the neutron signals. Expected sensitivities are -1g of fissile actinide and -100g of total actinide. A measure of spatial distribution is obtainable. The neutron time spectra obtained during neutron interrogation were more complex than expected, and more analysis is needed. Another area of discrepancy is the difference between predicted and measured thermal neutron flux in the drum. Clusters of small 3 He proportional counters were found to be much superior for fast neutron detection than larger diameter counters. It is necessary to ensure constancy of electron beam position relative to target(s) and drum, and prudent to measure the target neutron or gamma output as appropriate. 59 refs., 77 figs., 11 tabs

  3. A technique for combining neutron and gamma-ray data into a single assay value

    International Nuclear Information System (INIS)

    Pickrell, M.M.; Mercer, D.; Sharpe, T.J.

    1998-01-01

    The authors explored the potentials of using both neutron and gamma-ray measurements on a single item and combining these data into a single assay value. The purpose was to improve assay capability for sample matrices that are difficult to measure. They chose an empirical approach because they wanted to address difficult-to-measure items for which the assay problem is complex. They used the tomographic gamma scanner; a passive, high-efficiency neutron counter with add-a-source and multiplicity; and an active neutron, californium shuffler to obtain measurements. Twenty-four 200-L drums were measured with various matrices using all three machines. The matrices were chosen specifically to spain the difficult-to-measure assay problems for some or all of the instruments. For example, the authors measured a drum filled with concrete and another filled with metal. The data from these measurements were analyzed using the alternating conditional expectation algorithm, which is one of a class of generalized additive models. Other data fusion algorithms are also possible and are being explored. The intent was to find ways to combine the data that would reduce the matrix-induced measurement error

  4. Analysis and databasing software for integrated tomographic gamma scanner (TGS) and passive-active neutron (PAN) assay systems

    International Nuclear Information System (INIS)

    Estep, R.J.; Melton, S.G.; Buenafe, C.

    2000-01-01

    The CTEN-FIT program, written for Windows 9x/NT in C++,performs databasing and analysis of combined thermal/epithermal neutron (CTEN) passive and active neutron assay data and integrates that with isotopics results and gamma-ray data from methods such as tomographic gamma scanning (TGS). The binary database is reflected in a companion Excel database that allows extensive customization via Visual Basic for Applications macros. Automated analysis options make the analysis of the data transparent to the assay system operator. Various record browsers and information displays simplify record keeping tasks

  5. Non-destructive assay of mechanical components using gamma-rays and thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Erica Silvani; Avelino, Mila R. [PPG-EM/UERJ, R. Sao Francisco Xavier, 524, Maracana - Rio de Janeiro - RJ (Brazil); Almeida, Gevaldo L. de; Souza, Maria Ines S. [IEN/CNEN, Rua Helio de Almeida, 75, Ilha do Fundao, Rio de Janeiro - RJ (Brazil)

    2013-05-06

    This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermal neutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermal neutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

  6. 233U Assay A Neutron NDA System

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-11-17

    The assay of highly enriched {sup 233}U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched {sup 235}U do not convert easily over to the assay of {sup 233}U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with {gamma} ray isotopics information should give a good overall determination of {sup 233}U material now stored in bldg. 3019 at the Oak Ridge National Laboratory.

  7. Neutron and gamma-ray nondestructive examination of contact-handled transuranic waste at the ORNL TRU Waste Drum Assay Facility

    International Nuclear Information System (INIS)

    Schultz, F.J.; Coffey, D.E.; Norris, L.B.; Haff, K.W.

    1985-03-01

    A nondestructive assay system, which includes the Neutron Assay System (NAS) and the Segmented Gamma Scanner (SGS), for the quantification of contact-handled (<200 mrem/h total radiation dose rate at contact with container) transuranic elements (CH-TRU) in bulk solid waste contained in 208-L and 114-L drums has been in operation at the Oak Ridge National Laboratory since April 1982. The NAS has been developed and demonstrated by Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) for use by most US Department of Energy Defense Plant (DOE-DP) sites. More research and development is required, however, before the NAS can provide complete assay results for other than routine defense waste. To date, 525 ORNL waste drums have been assayed, with varying degrees of success. The isotopic complexity of the ORNL waste creates a correspondingly complex assay problem. The NAS and SGS assay data are presented and discussed. Neutron matrix effects, the destructive examination facility, and enriched uranium fuel-element assays are also discussed

  8. Use of delayed gamma rays for active non-destructive assay of {sup 235}U irradiated by pulsed neutron source (plasma focus)

    Energy Technology Data Exchange (ETDEWEB)

    Andola, Sanjay; Niranjan, Ram [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, T.C., E-mail: tckk@barc.gov.in [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Rout, R.K. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ashwani; Paranjape, D.B.; Kumar, Pradeep; Tomar, B.S.; Ramakumar, K.L. [Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, S.C. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-07-01

    A pulsed neutron source based on plasma focus device has been used for active interrogation and assay of {sup 235}U by monitoring its delayed high energy γ-rays. The method involves irradiation of fissile material by thermal neutrons obtained after moderation of a burst of neutrons emitted upon fusion of deuterium in plasma focus (PF) device. The delayed gamma rays emitted from the fissile material as a consequence of induced fission were detected by a large volume sodium iodide (NaI(Tl)) detector. The detector is coupled to a data acquisition system of 2k input size with 2k ADC conversion gain. Counting was carried out in pulse height analysis mode for time integrated counts up to 100 s while the temporal profile of delayed gamma has been obtained by counting in multichannel scaling mode with dwell time of 50 ms. To avoid the effect of passive (natural) and active (from surrounding materials) backgrounds, counts have been acquired for gamma energy between 3 and 10 MeV. The lower limit of detection of {sup 235}U in the oxide samples with this set-up is estimated to be 14 mg.

  9. Neutron Generators for Spent Fuel Assay

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard A.

    2010-01-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

  10. A simple neutron-gamma discriminating system

    International Nuclear Information System (INIS)

    Liu Zhongming; Xing Shilin; Wang Zhongmin

    1986-01-01

    A simple neutron-gamma discriminating system is described. A detector and a pulse shape discriminator are suitable for the neutron-gamma discriminating system. The influence of the constant fraction discriminator threshold energy on the neutron-gamma resolution properties is shown. The neutron-gamma timing distributions from an 241 Am-Be source, 2.5 MeV neutron beam and 14 MeV neutron beam are presented

  11. Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    Science.gov (United States)

    Jallu, F.; Loche, F.

    2008-08-01

    Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low α-activity fissile masses (mainly 235U, 239Pu, 241Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303-314]) based on ANI has been under development since 1996 to reach the incinerating α low level waste (LLW) criterion of about 50 Bq[α] per gram of crude waste (≈50 μg Pu) in 118 l drums on the date the drums are conditioned. Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm -3) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix ( d = 0.253 g cm -3). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix. The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction method, which consists in identifying and quantifying

  12. Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    International Nuclear Information System (INIS)

    Jallu, F.; Loche, F.

    2008-01-01

    Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low α-activity fissile masses (mainly 235 U, 239 Pu, 241 Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303-314]) based on ANI has been under development since 1996 to reach the incinerating α low level waste (LLW) criterion of about 50 Bq[α] per gram of crude waste (∼50 μg Pu) in 118 l drums on the date the drums are conditioned. Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm -3 ) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix (d = 0.253 g cm -3 ). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix. The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction method, which consists in identifying and

  13. Improvement of non-destructive fissile mass assays in {alpha} low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Jallu, F. [Commissariat a l' Energie Atomique, CEA, DEN, Nuclear Measurement Laboratory, Bat. 224, 13108 Saint Paul lez Durance (France)], E-mail: fanny.jallu@cea.fr; Loche, F. [Commissariat a l' Energie Atomique, CEA, DEN, Nuclear Measurement Laboratory, Bat. 224, 13108 Saint Paul lez Durance (France)

    2008-08-15

    Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low {alpha}-activity fissile masses (mainly {sup 235}U, {sup 239}Pu, {sup 241}Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303-314]) based on ANI has been under development since 1996 to reach the incinerating {alpha} low level waste (LLW) criterion of about 50 Bq[{alpha}] per gram of crude waste ({approx}50 {mu}g Pu) in 118 l drums on the date the drums are conditioned. Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm{sup -3}) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix (d = 0.253 g cm{sup -3}). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix. The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction

  14. Prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Goswami, A.

    2003-01-01

    Prompt gamma neutron activation analysis (PGNAA) is a technique for the analysis of elements present in solid, liquid and gaseous samples by measuring the capture gamma rays emitted from the sample during neutron irradiation. The technique is complementary to conventional neutron activation analysis (NAA) as it can be used in number of cases where NAA fails. Though the technique was first used in sixties, the advantage of the technique was first highlighted by Lindstrom and Anderson. PGNAA is increasingly being used as a rapid, instrumental, nondestructive and multielement analysis technique. A monograph and several excellent reviews on this topic have appeared recently. In this review, an attempt has been made to bring out the essential aspects of the technique, experimental arrangement and instrumentation involved, and areas of application. Some of the results will also be presented

  15. Analytical applications of neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Lindstrom, R.M.; Paul, R.L.; Anderson, D.L.; Paul, R.L.

    1997-01-01

    Field and industrial applications of neutron capture gamma-ray spectrometry with isotopic sources or neutron generators are economically important. Geochemical exploration in boreholes is done routinely with neutron probes. Coal and ores are assayed with analyzers adjacent to a conveyor belt in dozens of industrial facilities. The use of capture gamma rays for explosives detection has been described in the literature, both for scanning airline baggage and for characterizing obsolete munitions; a packaged system for the latter is available commercially. Generalizations are drawn from the history of the field, and predictions are made about the future usefulness of capture gamma rays. (author)

  16. Selection of non-destructive assay methods: Neutron counting or calorimetric assay?

    International Nuclear Information System (INIS)

    Cremers, T.L.; Wachter, J.R.

    1994-01-01

    The transition of DOE facilities from production to D ampersand D has lead to more measurements of product, waste, scrap, and other less attractive materials. Some of these materials are difficult to analyze by either neutron counting or calorimetric assay. To determine the most efficacious analysis method, variety of materials, impure salts and hydrofluorination residues have been assayed by both calorimetric assay and neutron counting. New data will be presented together with a review of published data. The precision and accuracy of these measurements are compared to chemistry values and are reported. The contribution of the gamma ray isotopic determination measurement to the overall error of the calorimetric assay or neutron assay is examined and discussed. Other factors affecting selection of the most appropriate non-destructive assay method are listed and considered

  17. Neutron detection gamma ray sensitivity criteria

    International Nuclear Information System (INIS)

    Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Mace, Emily K.; Stephens, Daniel L.; Woodring, Mitchell L.

    2011-01-01

    The shortage of 3 He has triggered the search for effective alternative neutron detection technologies for national security and safeguards applications. Any new detection technology must satisfy two basic criteria: (1) it must meet a neutron detection efficiency requirement, and (2) it must be insensitive to gamma-ray interference at a prescribed level, while still meeting the neutron detection requirement. It is the purpose of this paper to define measureable gamma ray sensitivity criteria for neutron detectors. Quantitative requirements are specified for: intrinsic gamma ray detection efficiency and gamma ray absolute rejection. The gamma absolute rejection ratio for neutrons (GARRn) is defined, and it is proposed that the requirement for neutron detection be 0.9 3 He based neutron detector is provided showing that this technology can meet the stated requirements. Results from tests of some alternative technologies are also reported.

  18. 233U Assay A Neutron NDA System

    International Nuclear Information System (INIS)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-01-01

    The assay of highly enriched 233 U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched 235 U do not convert easily over to the assay of 233 U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with γ ray isotopics information should give a good overall determination of 233 U material now stored in bldg. 3019 at the Oak Ridge National Laboratory

  19. Nondestructive assay of subassemblies of various spent or fresh fuels by active neutron interrogation

    International Nuclear Information System (INIS)

    Ragan, G.L.; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.

    1979-01-01

    Recent studies show that subassemblies containing various spent fuels could be assayed rapidly and accurately by a nondestructive assay system using active neutron interrogation and prompt-neutron detection. Subassembly penetration is achieved by 24-keV (Sb--Be) interrogation neutrons; the spent-fuel neutron background is overridden by using strong interrogating sources and prompt-neutron signals, and background gammas are absorbed by lead. Experiments have demonstrated the potential for assaying with better than 5% accuracy, three spent plutonium-fueled subassemblies per hour. Calculations, validated by experiments, predict even better performance for fresh or uranium-fueled subassemblies; several performance estimates are given

  20. Use of neutron-capture plastic fibers for nondestructive assay

    International Nuclear Information System (INIS)

    Heger, A.S.; Grazioso, R.F.; Mayo, D.R.; Ensslin, N.; Miller, M.C.; Huang, H.Y.; Russo, P.A.

    1998-01-01

    Neutron-capture plastic fibers can be used as a nondestructive assay tool. The detectors consist of an active region assembled from ribbons of boron-( 10 B) loaded optical fibers. The mixture of the moderator and thermal neutron absorber in the fiber yields a detector with high efficiency (var-epsilon) and a short die-away time (τ). The deposited energy of the resultant charged particles is converted to light that is collected by photomultiplier tubes mounted at both ends of the fiber. Thermal neutron coincidence counters (TNCC) made of these fibers can serve to verify fissile materials generated from the nuclear fuel cycle. This type of detector may extend the range of materials now accessible to assay by 3 He detectors. Experiments with single fibers of diameters 0.25, 0.50, and 1.00 mm test their ability to distinguish between the signals generated from neutron interactions and those from gamma rays. These results are compared with those obtained from simulation analyses for the same purpose. Light output and attenuation, neutron detection efficiency, and the signal-to-noise ratios of these fibers have also been investigated. The experimental results for light attenuation and neutron detection efficiency are consistent with the values obtained from simulation studies. A comparison of the performance of various configurations of the plastic scintillating fibers with that of other neutron-capture devices such as 3 He detectors is also discussed

  1. Simultaneous neutron and gamma spectrum adjustment

    International Nuclear Information System (INIS)

    Remec, I.

    1996-01-01

    The spectrum adjustment procedure was extended to simultaneous neutron and gamma spectrum adjustment, and the feasibility of this technique is demonstrated in the analysis of HFIR dosimetry experiments. Conditions in which gamma rays may contribute considerably to radiation damage in steels are discussed. Beryllium helium accumulation fluence monitors (HAFMs) were found to be good monitors in gamma fields of intensities high enough to contribute to steel embrittlement. Use of 237 Np, 238 U, and 9 Be HAFM as gamma dosimeters is proposed for high-dose irradiations in high-energy, high-intensity gamma fields

  2. The synchronous active neutron detection assay system

    International Nuclear Information System (INIS)

    Pickrell, M.M.; Kendall, P.K.

    1994-01-01

    We have begun to develop a novel technique for active neutron assay of fissile material in spent nuclear fuel. This approach will exploit a 14-MeV neutron generator developed by Schlumberger. The technique, termed synchronous active neutron detection (SAND), follows a method used routinely in other branches of physics to detect very small signals in presence of large backgrounds. Synchronous detection instruments are widely available commercially and are termed ''lock-in'' amplifiers. We have implemented a digital lock-in amplifier in conjunction with the Schlumberger neutron generator to explore the possibility of synchronous detection with active neutrons. The Schlumberger system can operate at up to a 50% duty factor, in effect, a square wave of neutron yield. Results are preliminary but promising. The system is capable of resolving the fissile material contained in a small fraction of the fuel rods in a cold fuel assembly; it also appears resilient to background neutron interference. The interrogating neutrons appear to be non-thermal and penetrating. Work remains to fully explore relevant physics and optimize instrument design

  3. A gamma-ray discriminating neutron scintillator

    International Nuclear Information System (INIS)

    Eschbach, P.A.; Miller, S.D.; Cole, M.C.

    1994-01-01

    A neutron scintillator has been developed at Pacific Northwest Laboratory which responds directly to as little as 10 mrem/hour dose equivalent rate fast neutron fields. The scintillator is composed of CaF 2 :Eu or of NaI grains within a silicone rubber or polystyrene matrix, respectively. Neutrons colliding with the plastic matrix provide knockon protons, which in turn deposit energy within the grains of phosphor to produce pulses of light. Neutron interactions are discriminated from gamma-ray events on the basis of pulse height. Unlike NE-213 liquid scintillators, this solid scintillator requires no pulseshape discrimination and therefore requires less hardware. Neutron events are anywhere from two to three times larger than the gamma-ray exposures are compared to 0.7 MeV gamma-ray exposures. The CaF 2 :Eu/silicone rubber scintillator is nearly optically transparent, and can be made into a very sizable detector (4 cm x 1.5 cm) without degrading pulse height. This CaF 2 :Eu scintillator has been observed to have an absolute efficiency of 0.1% when exposed to 5-MeV accelerator-generated neutrons (where the absolute efficiency is the ratio of observed neutron events divided by the number of fast neutrons striking the detector)

  4. Plasma driven neutron/gamma generator

    Science.gov (United States)

    Leung, Ka-Ngo; Antolak, Arlyn

    2015-03-03

    An apparatus for the generation of neutron/gamma rays is described including a chamber which defines an ion source, said apparatus including an RF antenna positioned outside of or within the chamber. Positioned within the chamber is a target material. One or more sets of confining magnets are also provided to create a cross B magnetic field directly above the target. To generate neutrons/gamma rays, the appropriate source gas is first introduced into the chamber, the RF antenna energized and a plasma formed. A series of high voltage pulses are then applied to the target. A plasma sheath, which serves as an accelerating gap, is formed upon application of the high voltage pulse to the target. Depending upon the selected combination of source gas and target material, either neutrons or gamma rays are generated, which may be used for cargo inspection, and the like.

  5. Neutron and gamma-ray toxicity studies

    International Nuclear Information System (INIS)

    Ainsworth, E.J.

    1975-01-01

    The focus of the program is on late effects of neutron and gamma radiation and assessment of risk. Principal research activities are in two complementary areas: life-span experiments with large populations of laboratory mice to compare the effectiveness of single or protracted doses of neutron or gamma radiation for life shortening due to cancer and other debilitating noncancerous diseases; and basic research on cellular injury and recovery for the evaluation of potential contributions of latent injury in the mouse circulatory, immune, and hematopoietic systems to life shortening, and for the comparison of late radiation effects in proliferating tissues. The data are used to test existing models and to formulate new models for prediction of radiation hazards and the relative biological effectiveness (RBE) of fission neutrons, particularly at low radiation doses. The neutron dose-response curve is nonlinear, with the life shortening effect decreasing from 3-4 day/rad to 1 day/rad with increasing dose over the range of 20-240 rad. Clearly, linear extrapolations from high neutron doses to estimate life shortening at low doses would underestimate risk; the underestimation is even greater when the enhancement of life shortening produced by fractionated neutron exposure, described previously by us, is also considered. These results from single neutron doses deviate from predictions of total dose dependency based on the predictive model of Kellerer and Rossi. The shape of the gamma radiation dose-response curve is linear over the range of 90 to 788 rad; linear dose-response curves for gamma radiation have been described previously by others, but a quadratic function has been considered by some to be most applicable

  6. Neutron gamma competition in fast fission

    International Nuclear Information System (INIS)

    Frehaut, J.

    1989-01-01

    In the present paper we analyse the data we have obtained on the distribution of the gamma-ray energy per fission, as well as on the average energy E-barγ released per fission for the neutron induced fission of several isotopes, in the energy range up to 15 MeV. 6 refs, 9 figs

  7. Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources

    CERN Document Server

    Matsumoto, T

    2003-01-01

    Two small neutron sources of sup 2 sup 5 sup 2 Cf and sup 2 sup 4 sup 1 Am-Be radioisotopes were used for design of neutron beams applicable to low intensity neutron and gamma ray radioscopy (LINGR). In the design, Monte Carlo code (MCNP) was employed to generate neutron and gamma ray beams suited to LINGR. With a view to variable neutron spectrum and neutron intensity, various arrangements were first examined, and neutron-filter, gamma-ray shield and beam collimator were verified. Monte Carlo calculations indicated that with a suitable filter-shield-collimator arrangement, thermal neutron beam of 3,900 ncm sup - sup 2 s sup - sup 1 with neutron/gamma ratio of 7x10 sup 7 , and 25 ncm sup - sup 2 s sup - sup 1 with very large neutron/gamma ratio, respectively, could be produced by using sup 2 sup 5 sup 2 Cf(122 mu g) and a sup 2 sup 4 sup 1 Am-Be(37GBq)radioisotopes at the irradiation port of 35 cm from the neutron sources.

  8. Neutron-gamma discrimination of boron loaded plastic scintillator

    International Nuclear Information System (INIS)

    Wang Dong; He Bin; Zhang Quanhu; Wu Chuangxin; Luo Zhonghui

    2010-01-01

    Boron loaded plastic scintillator could detect both fast neutrons thanks to hydrogen and thermal neutrons thanks to 10B. Both reactions have large cross sections, and results in high detection efficiency of incident neutrons. However, similar with other organic scintillators, boron loaded plastic scintillator is sensitive to gamma rays and neutrons. So gamma rays must be rejected from neutrons using their different behavior in the scintillator. In the present research zero crossing method was used to test neutron-gamma discrimination of BC454 boron loaded plastic scintillator. There are three Gaussian peaks in the time spectrum, they are corresponding to gamma rays, fast neutrons and flow neutrons respectively. Conclusion could be made that BC454 could clear discriminate slow neutrons and gamma, but the discrimination performance turns poor as the neutrons' energy becomes larger. (authors)

  9. Thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Tuli, J.K.

    1983-01-01

    The energy and intensity of gamma rays as seen in thermal neutron capture are presented. Only those (n,α), E = thermal, reactions for which the residual nucleus mass number is greater than or equal to 45 are included. These correspond to evaluations published in Nuclear Data Sheets. The publication source data are contained in the Evaluated Nuclear Structure Data File (ENSDF). The data presented here do not involve any additional evaluation. Appendix I lists all the residual nuclides for which the data are included here. Appendix II gives a cumulated index to A-chain evaluations including the year of publication. The capture gamma ray data are given in two tables - the Table 1 is the list of all gamma rays seen in (n,#betta#) reaction given in the order of increasing energy; the Table II lists the gamma rays according to the nuclide

  10. Design innovations in neutron and gamma detectors

    International Nuclear Information System (INIS)

    Prasad, K.R.

    2003-01-01

    Neutron and gamma radiation needs to be monitored in most nuclear installations since it is highly penetrating. On-line monitoring of these radiations is very important for the safe and controlled operation of nuclear reactors, accelerators etc. Several design innovations have been carried out on gas ionisation detectors such as boron-lined proportional counters and ion chambers, fission detectors, gamma ion chambers as well as self-powered detectors. The use of additional structures within boron-lined detectors has enhanced their neutron sensitivity without a corresponding increase in the unwanted gamma sensitivity. The neutron sensitivity of fission counters can be enhanced by designing them as transmission line devices. Ion chambers with two and six pairs of electrodes have been developed for monitoring pulsed x-ray background at accelerator areas. Ion chambers have been employed at gamma fields up to 80 kR/h by deriving the exposure levels on-line using microcontroller devices programmed on the basis of theoretical and empirical formulas. The use of gas electron multiplier foils is proposed for charge multiplication in ion chambers. Self-powered detectors with new emitter materials like Hi, Ni and Inconel have been developed. (author)

  11. High Energy Neutron Induced Gamma Production

    International Nuclear Information System (INIS)

    Brown, D.A.; Johnson, M.; Navratil, P.

    2007-01-01

    N Division has an interest in improving the physics and accuracy of the gamma data it provides to its customers. It was asked to look into major gamma producing reactions for 14 MeV incident neutrons for several low-Z materials and determine whether LLNL's processed data files faithfully represent the current state of experimental and theoretical knowledge for these reactions. To address this, we surveyed the evaluations of the requested materials, made recommendations for the next ENDL release and noted isotopes that will require further experimental study. This process uncovered several major problems in our translation and processing of the ENDF formatted evaluations, most of which have been resolved

  12. Bulk media assay using backscattered neutron spectrometry

    International Nuclear Information System (INIS)

    Csikai, J.

    2000-01-01

    This paper summarized a systematic study of bulk media assay using backscattered neutron spectrometry. The source-sample-detector geometry used for the measurements of leakage and elastically backscattered (EBS) spectra of neutrons is shown. Neutrons up to about 14 MeV were produced via 2 H (d,n) and 9 Be (d,n) reactions using different deuteron beam energies between 5 and 10 MeV at the MGC-20E cyclotron of ATOMKI (Debrecen). Neutron yields of the Pu-Be and 252 Cf sources were 5.25 x 10 6 n/s and 1.8 x 10 6 n/s, respectively. Flux density distributions of thermal and primary 14 MeV neutrons were measured for graphite, water and coal samples in various moderator (M)-sample (S)-reflector (R) geometries. Relative fractions and integrated yields of 252 Cf, Pu-Be and 14 MeV neutrons above the (n,n'γ) reaction thresholds for 12 C, 16 O and 28 Si isotopes vs sample thickness have also been determined. It was found that the integrated reaction rate vs sample thickness decreasing exponentially with different attenuation coefficients depending on the neutron spectrum and the composition of the sample. The spectra of neutrons from sources passing through slabs of water, graphite, sand, Al, Fe and Pb up to 20 cm in thickness have been measured by a PHRS system in the 1.2 to 1.5 MeV range. The leakage neutron spectra from a Pu-Be source placed in the center of 30 cm diameter sphere filled with water, paraffin oil, SiO 2 , zeolite and river sand were also measured. The measured spectra have been compared with the calculated results obtained by the three dimensional Monte-Carlo code MCNP-4A and point-wise cross sections from the ENDF/B-4, ENDF/B-6, ENDF/E-1, BROND-2 and JENDL-3.1 data files. New results were obtained for validation of different data libraries from a comparison on the measured and the calculated spectra. Some typical results for water, Al, sand and Fe are shown. A combination of the backscattered neutron spectrometry with the surface gauge used both for the

  13. Discrimination methods between neutron and gamma rays for boron loaded plastic scintillators

    CERN Document Server

    Normand, S; Haan, S; Louvel, M

    2002-01-01

    Boron loaded plastic scintillators exhibit interesting properties for neutron detection in nuclear waste management and especially in investigating the amount of fissile materials when enclosed in waste containers. Combining a high thermal neutron efficiency and a low mean neutron lifetime, they are suitable in neutron multiplicity counting. However, due to their high sensitivity to gamma rays, pulse shape discrimination methods need to be developed in order to optimize the passive neutron assay measurement. From the knowledge of their physical properties, it is possible to separate the three kinds of particles that have interacted in the boron loaded plastic scintillator (gamma, fast neutron and thermal neutron). For this purpose, we have developed and compared the two well known discrimination methods (zero crossing and charge comparison) applied for the first time to boron loaded plastic scintillator. The setup for the zero crossing discrimination method and the charge comparison methods is thoroughly expl...

  14. Assay of plutonium contaminated waste by gamma spectrometry

    International Nuclear Information System (INIS)

    Adsley, I.; Bull, R.; Davies, M.; Green, M.

    2011-01-01

    The extreme toxicity of plutonium necessitates the segregation of plutonium contaminated materials (PCM) with extremely small (sub-μg) levels of contamination. The driver to measure accurately these small quantities of plutonium within (relatively) large volumes of waste is (in part) financial. In particular the cost of disposal (per unit volume) rises steeply with increasing waste-category. Within the UK, there has been a historical reluctance to use low energy gamma radiation to sentence PCM because of the potential for self attenuation by dense materials. This is unfortunate because the low-energy gamma radiation from PCM offers the only practicable technique for segregating PCM within the various Low Level Waste (LLW) (>0.4Bq/g) and sub-LLW categories. Whilst passive neutron counting techniques have proved successful for assay of waste well into the Intermediate Level Waste (ILW) (>100Bq/g) category, a cursory study reveals that these techniques are barely capable of detecting mg quantities of plutonium -- let alone the sub-μg quantities present in LLW. This paper considers the use of two types of gamma detector for assay of PCM: the thin sodium iodide FIDLER (Field Instrument for the Detection of Low Energy Radiation) and the HPGe (High Purity Germanium) detector. Systems utilising these two types of detector can provide complementary information. FIDLER measurements are conducted by careful, local, systematic monitoring of surfaces. By contrast a HPGe detector can be used to monitor entire walls, or even rooms, in one measurement. Thus, a HPGe detector placed in the centre of room (from which any radioactive hot-spots have previously been removed) could be used to demonstrate that the average activity remaining close to the surface of the walls/floor/ceiling is below a given limit. The Monte Carlo Code MCNP 1 has been used to model both FIDLER probe and HPGe detector in the measurement geometries described above. The MCNP simulations have been validated

  15. Virtual Gamma Ray Radiation Sources through Neutron Radiative Capture

    Energy Technology Data Exchange (ETDEWEB)

    Scott Wilde, Raymond Keegan

    2008-07-01

    The countrate response of a gamma spectrometry system from a neutron radiation source behind a plane of moderating material doped with a nuclide of a large radiative neutron capture cross-section exhibits a countrate response analogous to a gamma radiation source at the same position from the detector. Using a planar, surface area of the neutron moderating material exposed to the neutron radiation produces a larger area under the prompt gamma ray peak in the detector than a smaller area of dimensions relative to the active volume of the gamma detection system.

  16. Non-destructive assay of 242Pu by resonance neutron capture

    International Nuclear Information System (INIS)

    Kane, W.R.; Lu, Ming-Shih; Aronson, A.; Forman, L.; Vanier, P.E.

    1995-01-01

    For the accurate assay of plutonium by neutron correlation measurements, especially for material derived from high-burnup reactor fuel, the content of 242 Pu in a sample must be determined. Since 242 Pu has a long half-life (387,000 yr) and decays to 238 U by alpha particle emission with the accompanying emission of only weak, low-energy gamma rays, gamma-ray spectrometry methods which are ordinarily employed to determine the isotopic composition of a plutonium sample are not feasible for 242 Pu. The existence of a resonance in the neutron capture cross section of 242 Pu at an energy of 2.67 electron volts (eV) with a large (72, 000 barn) cross section affords the possibility for the quantitative assay of this isotope by epithermal neutron capture. Essential for this purpose is an appropriately designed geometry of neutron moderators and absorbers which will provide maximum flux in the eV region while suppressing thermal neutron capture by the fissile plutonium isotopes. Signatures for neutron capture in 242 Pu include the decay of 243 Pu (4.9 hr), prompt capture gamma rays (total energy 5.034 MeV), and the decay of an isomeric state (330 nanosecond). Experiments to determine the feasibility of this approach are currently in progress

  17. Earth formation pulsed neutron porosity logging system utilizing epithermal neutron and inelastic scattering gamma ray detectors

    International Nuclear Information System (INIS)

    Smith, H.D. Jr.; Smith, M.P.; Schultz, W.E.

    1978-01-01

    An improved pulsed neutron porosity logging system is provided in the present invention. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector and an inelastic scattering gamma ray detector is moved through a borehole. The detection of inelastic gamma rays provides a measure of the fast neutron population in the vicinity of the detector. repetitive bursts of neutrons irradiate the earth formation and, during the busts, inelastic gamma rays representative of the fast neutron population is sampled. During the interval between bursts the epithermal neutron population is sampled along with background gamma radiation due to lingering thermal neutrons. the fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity

  18. MCNP modelling of a combined neutron/gamma counter

    CERN Document Server

    Bourva, L C A; Ottmar, H; Weaver, D R

    1999-01-01

    A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...

  19. A Compton Suppressed Gamma Ray Counter For Radio Assay of Materials

    Science.gov (United States)

    Godfrey, Benjamin

    2016-03-01

    Rare event searches, such as direct dark matter experiments, require materials with ultra-low levels of natural radioactivity. We present a neutron activation analysis (NAA) technique for assaying metals, specifically titanium used for cryostat construction. Earlier attempts at NAA encountered limitations due to bulk activation via (n, p) reactions, which contributed to large continuum backgrounds due to Compton tails. Our method involves a heavy water shielded exposure to minimize (n,p) reactions and a sodium iodide shielded high purity germanium counter for the gamma ray assay. Preliminary results on assays for U/Th/K contamination in titaniumwill be presented.

  20. Biological dosimetry for mixed gamma-neutron field

    International Nuclear Information System (INIS)

    Brandao, J.O.C.; Santos, J.A.L.; Souza, P.L.G.; Lima, F.F.; Vilela, E.C.; Calixto, M.S.; Santos, N.

    2011-01-01

    There is increasing concern about airline crew members (about one million worldwide) exposed to measurable neutrons doses. Historically, cytogenetic biodosimetry assays have been based on quantifying asymmetrical chromosome alterations (dicentrics, centric rings and acentric fragments) in mitogen-stimulated T-lymphocytes in their first mitosis after radiation exposure. Increased levels of chromosome damage in peripheral blood lymphocytes are a sensitive indicator of radiation exposure and they are routinely exploited for assessing radiation absorbed dose after accidental or occupational exposure. Since radiological accidents are not common, not all nations feel that it is economically justified to maintain biodosimetry competence. However, dependable access to biological dosimetry capabilities is completely critical in event of an accident. In this paper the dose-response curve was measured for the induction of chromosomal alterations in peripheral blood lymphocytes after chronic exposure in vitro to mixed gamma-neutron field. Blood was obtained from one healthy donor and exposed to two mixed gamma-neutron field from sources 241 AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL - CRCN/NE - PE - Brazil). The evaluated absorbed doses were 0.2 Gy; 1.0 Gy and 2.5 Gy. The dicentric chromosomes were observed at metaphase, following colcemide accumulation and 1000 well-spread metaphases were analyzed for the presence of dicentrics by two experts after painted by giemsa 5%. The preliminary results showed a linear dependence between radiations absorbed dose and dicentric chromosomes frequencies. Dose-response curve described in this paper will contribute to the construction of calibration curve that will be used in our laboratory for biological dosimetry. (author)

  1. Assay of low-enriched uranium using spontaneous fission neutrons

    International Nuclear Information System (INIS)

    Zucker, M.S.; Fainberg, A.

    1980-01-01

    Low-enriched uranium oxide in bulk containers can be assayed for safeguards purposes, using the neutrons from spontaneous fission of 238 U as a signature, to complement enrichment and mass measurement. The penetrability of the fast fission neutrons allows the inner portion of bulk samples to register. The measurement may also be useful for measuring moisture content, of significance in process control. The apparatus used can be the same as for neutron correlation counting for Pu assay. The neutron multiplication observed in 238 U is of intrinsic interest

  2. Portable neutron and gamma-radiation instruments

    International Nuclear Information System (INIS)

    Murray, W.S.; Butterfield, K.B.

    1990-01-01

    This paper reports on the design and building of a smart neutron and gamma-radiation detection systems with embedded microprocessors programmed in the FORTH language. These portable instruments can be battery-powered and can provide many analysis functions not available in most radiation detectors. Local operation of the instruments is menu-driven through a graphics liquid crystal display and hex keypad; remote operation is through a serial communications link. While some instruments simply count particles, others determine the energy of the radiation as well as the intensity. The functions the authors have provided include absolute source-strength determination. Feynmann variance analysis, sequential-probability ratio test, and time-history recording

  3. Neutron and gamma-ray toxicity studies

    International Nuclear Information System (INIS)

    Ainsworth, E.J.

    1975-01-01

    Results are reported from studies on the late effects of irradiation on large populations of mice. The effectiveness of neutron and gamma radiation for production of neoplastic and non-neoplastic diseases and life shortening is compared. Basic studies of cellular and functional indices of radiation injury, which provide the opportunity for fundamental new contributions to the understanding of late radiation effects in the vascular, immune, and hematopoietic systems are also reported. Both structural and functional changes in the vasculature have been observed during the second year after irradiation. The structural changes in the pinna include collapse of arteries, arterioles, and some veins along with alterations in the smooth musculature and accumulation of significant fibrosis. Late ultrastructural changes observed in myofibrils involve the endoplasmic reticulum and mitochondria. Cardiac muscle also showed alteration in the size and number of mitochondria, and fibrosis development within 7 days of irradiation. (U.S.)

  4. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  5. Neutron-gamma discrimination by pulse analysis with superheated drop detector

    International Nuclear Information System (INIS)

    Das, Mala; Seth, S.; Saha, S.; Bhattacharya, S.; Bhattacharjee, P.

    2010-01-01

    Superheated drop detector (SDD) consisting of drops of superheated liquid of halocarbon is irradiated to neutrons and gamma-rays from 252 Cf fission neutron source and 137 Cs gamma source, respectively, separately. Analysis of pulse height of signals at the neutron and gamma-ray sensitive temperature provides significant information on the identification of neutron and gamma-ray induced events.

  6. Estimation of neutron energy distributions from prompt gamma emissions

    Science.gov (United States)

    Panikkath, Priyada; Udupi, Ashwini; Sarkar, P. K.

    2017-11-01

    A technique of estimating the incident neutron energy distribution from emitted prompt gamma intensities from a system exposed to neutrons is presented. The emitted prompt gamma intensities or the measured photo peaks in a gamma detector are related to the incident neutron energy distribution through a convolution of the response of the system generating the prompt gammas to mono-energetic neutrons. Presently, the system studied is a cylinder of high density polyethylene (HDPE) placed inside another cylinder of borated HDPE (BHDPE) having an outer Pb-cover and exposed to neutrons. The emitted five prompt gamma peaks from hydrogen, boron, carbon and lead can be utilized to unfold the incident neutron energy distribution as an under-determined deconvolution problem. Such an under-determined set of equations are solved using the genetic algorithm based Monte Carlo de-convolution code GAMCD. Feasibility of the proposed technique is demonstrated theoretically using the Monte Carlo calculated response matrix and intensities of emitted prompt gammas from the Pb-covered BHDPE-HDPE system in the case of several incident neutron spectra spanning different energy ranges.

  7. Standardization of portable assay instrumentation: the neutron-coincidence tree

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1983-01-01

    Standardization of portable neutron assay instrumentation has been achieved by using the neutron coincidence technique as a common basis for a wide range of instruments and applications. The electronics originally developed for the High-Level Neutron Coincidence Counter has been adapted to both passive- and active-assay instrumentation for field verification of bulk plutonium, inventory samples, pellets, powders, nitrates, high-enriched uranium, and materials-testing-reactor, light-water-reactor, and mixed-oxide fuel assemblies. The family of detectors developed at Los Alamos National Laboratory and their performance under in-field conditions are described. 16 figures, 3 tables

  8. Neutron and gamma irradiation effects on power semiconductor switches

    Science.gov (United States)

    Schwarze, G. E.; Frasca, A. J.

    1990-01-01

    The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

  9. Neutron and gamma irradiation damage to organic materials.

    Energy Technology Data Exchange (ETDEWEB)

    White, Gregory Von, II; Bernstein, Robert

    2012-04-01

    This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

  10. The Passive Neutron Enrichment Meter for Uranium Cylinder Assay

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Karen A.; Menlove, Howard O.; Swinhoe, Martyn T.; Marlow, Johanna B. [Safeguards Science and Technology Group (N-1), Los Alamos National Laboratory, Los Alamos (United States)

    2011-12-15

    As fuel cycle technology becomes more prevalent around the world, international safeguards have become increasingly important in verifying that nuclear materials have not been diverted. Uranium enrichment technology is a critical pathway to nuclear weapons development, making safeguards of enrichment facilities especially important. Independently-verifiable material accountancy is a fundamental measure in detecting diversion of nuclear materials. This paper is about a new instrument for uranium cylinder assay for enrichment plant safeguards called the Passive Neutron Enrichment Meter (PNEM). The measurement objective is to simultaneously verify uranium mass and enrichment in Uf6 cylinders. It can be used with feed, product, and tails cylinders. Here, we consider the enrichment range up to 5% {sup 235}U. The concept is to use the Doubles-to-Singles count rate to give a measure of the {sup 235}U enrichment and the Singles count rate to provide a measure of the total uranium mass. The cadmium ratio is an additional signature for the enrichment that is especially useful for feed and tails cylinders. PNEM is a {sup 3}He-based system that consists of two portable detector pods. Uranium enrichment in UF{sub 6} cylinders is typically determined using a gamma-ray-based method that only samples a tiny volume of the cylinder's content and requires knowledge of the cylinder wall thickness. The PNEM approach has several advantages over gamma-ray-based methods including a deeper penetration depth into the cylinder, meaning it can be used with heterogeneous isotopic mixtures of UF{sub 6}. In this paper, we describe a Monte Carlo modelling study where we have examined the sensitivity of the system to systematic uncertainties such as the distribution of UF{sub 6} within the cylinder. We also compare characterization measurements of the PNEM prototype to the expected measurements calculated with Monte Carlo simulations.

  11. Comparison of gamma, neutron and proton irradiations of multimode fibers

    International Nuclear Information System (INIS)

    Gingerich, M.E.; Dorsey, K.L.; Askins, C.G.; Friebele, E.J.

    1987-01-01

    The effects of pure gamma, pure proton, and mixed neutron-gamma irradiation fields on a set of both pure and doped silica core multimode fibers have been investigated. Only slight differences are found in the radiation response of pure and doped silica core fibers exposed to gamma or mixed neutron-gamma fields, indicating that Co-60 sources can be used to simulate the effects of the mixed field (except in the case of a pure neutron environment). Although it is noted that neither mix field nor gamma sources adequately simulate the effects of proton irradiation of doped silica core fibers, a good correspondence is found in the case of the pure silica core waveguide. 13 references

  12. Formulation of the relationship between indices of neutron-gamma and gamma-gamma method and the percentrage of iron

    International Nuclear Information System (INIS)

    Majorowicz, J.

    1973-01-01

    In this article, the author presents the possibility of a complex utilization of radiometric logging methods, neutron-gamma profiling and gamma-gamma density logging for determining percentage of iron and establishing geophysical possibilities of identifying zones of economically profitable ores in borehole profiles. Figures present the correlations between indices of neutron-gamma and gamma-gamma logging methods and the percentage of iron, as well as the correlation of neutron-gamma and gamma-gamma indices for zones minerallized with iron ores. The article presents the correlational analyses of the results: the correlational coefficients are given as well as total error in determining iron content on the basis of each of the methods described. Next, a multidimensional statistical analysis is carried out on the results obtained. On the basis of the two-dimensional correlational coefficients calculated and the average standard deviation, an equation of linear regression was formulated, simultaneously involving three parameters - the indices of neutron-gamma and gamma-gamma logging and the percentage of iron. The multiple correlational coefficient obtained markedly exceeds the two-dimentional correlation coefficient (r=0.974>rsub(xz)>rsub(yz)>rsub(xy)). The given method of utilizing multidimensional statistics in borehole geophysics for identifying iron ores is an efficient one. On the basis of several relationships among independent variables which are less obvious (smaller values of correlational coefficient), it is possible to obtain a single distinct relationship involving all variables simultaneously. (author)

  13. Self-powered neutron and gamma-ray flux detector

    International Nuclear Information System (INIS)

    Allan, C.J.; Shields, R.B.; Lynch, G.F.; Cuttler, J.M.

    1980-01-01

    A new type of self-powered neutron detector was developed which is sensitive to both the neutron and gamma-ray fluxes. The emitter comprises two parts. The central emitter core is made of materials that generate high-energy electrons on exposure to neutrons. The outer layer acts as a gamma-ray/electron converter, and since it has a higher atomic number and higher back-scattering coefficient than the collector, increases the net outflow or emmission of electrons. The collector, which is around the emitter outer layer, is insulated from the outer layer electrically with dielectric insulation formed from compressed metal-oxide powder. The fraction of electrons given off by the emitter that is reflected back by the collector is less than the fraction of electrons emitted by the collector that is reflected back by the emitter. The thickness of the outer layer needed to achieve this result is very small. A detector of this design responds to external reactor gamma-rays as well as to neutron capture gamma-rays from the collector. The emitter core is either nickel, iron or titanium, or alloys based on these metals. The outer layer is made of platinum, tantalum, osmium, molybdenum or cerium. The detector is particularly useful for monitoring neutron and gamma ray flux intensities in nuclear reactor cores in which the neutron and gamma ray flux intensities are closely proportional, are unltimately related to the fission rate, and are used as measurements of nuclear reactor power. (DN)

  14. Some target assay uncertainties for passive neutron coincidence counting

    International Nuclear Information System (INIS)

    Ensslin, N.; Langner, D.G.; Menlove, H.O.; Miller, M.C.; Russo, P.A.

    1990-01-01

    This paper provides some target assay uncertainties for passive neutron coincidence counting of plutonium metal, oxide, mixed oxide, and scrap and waste. The target values are based in part on past user experience and in part on the estimated results from new coincidence counting techniques that are under development. The paper summarizes assay error sources and the new coincidence techniques, and recommends the technique that is likely to yield the lowest assay uncertainty for a given material type. These target assay uncertainties are intended to be useful for NDA instrument selection and assay variance propagation studies for both new and existing facilities. 14 refs., 3 tabs

  15. A novel dual mode neutron-gamma imager

    International Nuclear Information System (INIS)

    Cooper, Robert Lee; Gerling, Mark; Brennan, James S.; Mascarenhas, Nicholas; Mrowka, Stanley; Marleau, Peter

    2010-01-01

    The Neutron Scatter Camera (NSC) can image fission sources and determine their energy spectra at distances of tens of meters and through significant thicknesses of intervening materials in relatively short times (1). We recently completed a 32 element scatter camera and will present recent advances made with this instrument. A novel capability for the scatter camera is dual mode imaging. In normal neutron imaging mode we identify and image neutron events using pulse shape discrimination (PSD) and time of flight in liquid scintillator. Similarly gamma rays are identified from Compton scatter in the front and rear planes for our segmented detector. Rather than reject these events, we show it is possible to construct a gamma-ray image by running the analysis in a 'Compton mode'. Instead of calculating the scattering angle by the kinematics of elastic scatters as is appropriate for neutron events, it can be found by the kinematics of Compton scatters. Our scatter camera has not been optimized as a Compton gamma-ray imager but is found to work reasonably. We studied imaging performance using a Cs137 source. We find that we are able to image the gamma source with reasonable fidelity. We are able to determine gamma energy after some reasonable assumptions. We will detail the various algorithms we have developed for gamma image reconstruction. We will outline areas for improvement, include additional results and compare neutron and gamma mode imaging.

  16. A method to describe inelastic gamma field distribution in neutron gamma density logging.

    Science.gov (United States)

    Zhang, Feng; Zhang, Quanying; Liu, Juntao; Wang, Xinguang; Wu, He; Jia, Wenbao; Ti, Yongzhou; Qiu, Fei; Zhang, Xiaoyang

    2017-11-01

    Pulsed neutron gamma density logging (NGD) is of great significance for radioprotection and density measurement in LWD, however, the current methods have difficulty in quantitative calculation and single factor analysis for the inelastic gamma field distribution. In order to clarify the NGD mechanism, a new method is developed to describe the inelastic gamma field distribution. Based on the fast-neutron scattering and gamma attenuation, the inelastic gamma field distribution is characterized by the inelastic scattering cross section, fast-neutron scattering free path, formation density and other parameters. And the contribution of formation parameters on the field distribution is quantitatively analyzed. The results shows the contribution of density attenuation is opposite to that of inelastic scattering cross section and fast-neutron scattering free path. And as the detector-spacing increases, the density attenuation gradually plays a dominant role in the gamma field distribution, which means large detector-spacing is more favorable for the density measurement. Besides, the relationship of density sensitivity and detector spacing was studied according to this gamma field distribution, therefore, the spacing of near and far gamma ray detector is determined. The research provides theoretical guidance for the tool parameter design and density determination of pulsed neutron gamma density logging technique. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Kalman filter analysis of delayed neutron nondestructive assay measurements

    International Nuclear Information System (INIS)

    Aumeier, S. E.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile and fertile nuclei in various matrices is important in several nuclear applications including international and domestics safeguards, radioactive waste characterization and nuclear facility operations. Material irradiation followed by delayed neutron counting is a well known and useful nondestructive assay technique used to determine the fissile-effective content of assay samples. Previous studies have demonstrated the feasibility of using Kalman filters to unfold individual isotopic contributions to delayed neutron measurements resulting from the assay of mixes of uranium and plutonium isotopes. However, the studies in question used simulated measurement data and idealized parameters. We present the results of the Kalman filter analysis of several measurements of U/Pu mixes taken using Argonne National Laboratory's delayed neutron nondestructive assay device. The results demonstrate the use of Kalman filters as a signal processing tool to determine the fissile and fertile isotopic content of an assay sample from the aggregate delayed neutron response following neutron irradiation

  18. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B. W.; Summers, N.; Escher, J.; Firestone, R. B.; Basunia, S.; Hurst, A.; Krticka, M.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

    2011-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

  19. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B.W.; Firestone, R.B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H.D.

    2010-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. this can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. They are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

  20. Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

    Science.gov (United States)

    Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

    2018-03-01

    In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

  1. Preliminary results of a neutron-gamma coincidence experiment

    International Nuclear Information System (INIS)

    Piercey, R.B.; Dunnam, F.E.; Muga, M.L.; Rester, A.C.; Ramayya, A.V.; Hamilton, J.H.; Eberth, J.; Zganjar, E.F.

    1984-01-01

    The recently completed neutron multiplicity detector dubbed PANDA (Pentagonal Annular Neutron Detector Array) is fully described later in this report. The new detector was recently used for the first time on-line at the Holifield Heavy Ion Research Facility to measure neutron-gamma coincidence in the 24 Mg( 58 Ni,xαypzn) reaction. The detector configuration for the experiment is shown. The PANDA was situated in the forward direction, coaxial to the beam line with five gamma-ray detectors placed at +/- 90 0 , +/- 135 0 , and 0 0 . 2 figures

  2. Real‑time, fast neutron detection for stimulated safeguards assay

    International Nuclear Information System (INIS)

    Joyce, Malcolm J.; Adamczyk, Justyna; Plenteda, Romano; Aspinall, Michael D.; Cave, Francis D.

    2015-01-01

    The advent of low‑hazard organic liquid scintillation detectors and real‑time pulse‑shape discrimination (PSD) processing has suggested a variety of modalities by which fast neutrons, as opposed to neutrons moderated prior to detection, can be used directly to benefit safeguards needs. In this paper we describe a development of a fast‑neutron based safeguards assay system designed for the assessment of 235 U content in fresh fuel. The system benefits from real‑time pulse‑shape discrimination processing and auto‑calibration of the detector system parameters to ensure a rapid and effective set‑up protocol. These requirements are essential in optimising the speed and limit of detection of the fast neutron technique, whilst minimising the intervention needed to perform the assay.

  3. Neutron and gamma-ray transport experiments in liquid air

    International Nuclear Information System (INIS)

    Farley, W.E.

    1976-01-01

    Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent

  4. Physical principles of neutron-gamma materials monitoring

    Science.gov (United States)

    Pekarskii, G. Sh.

    1986-03-01

    The physical principles of secondary radiation methods in nondestructive testing are discussed. Among the techniques considered are: neutron activation analysis (NAA); the induced-radiation method; and quasialbedo recording of secondary gamma-radiation. Emphasis is given to the neutron-gamma method which consists of exposing test material to a neutron flux and recording the secondary gamma-radiation by means of a spectrometer. The limitations of the method in detecting local inhomogeneous defects (filled pores cracks, and inclusions) in metal layers and multicomponents materials are described, and some advantages of the method over NAA are discussed. Formulas are derived for estimating the optimum density of the gamma-ray flux which is received by the detector.

  5. Neutron counting and gamma spectroscopy with PVT detectors

    International Nuclear Information System (INIS)

    Mitchell, Dean James; Brusseau, Charles A.

    2011-01-01

    Radiation portals normally incorporate a dedicated neutron counter and a gamma-ray detector with at least some spectroscopic capability. This paper describes the design and presents characterization data for a detection system called PVT-NG, which uses large polyvinyl toluene (PVT) detectors to monitor both types of radiation. The detector material is surrounded by polyvinyl chloride (PVC), which emits high-energy gamma rays following neutron capture reactions. Assessments based on high-energy gamma rays are well suited for the detection of neutron sources, particularly in border security applications, because few isotopes in the normal stream of commerce have significant gamma ray yields above 3 MeV. Therefore, an increased count rate for high-energy gamma rays is a strong indicator for the presence of a neutron source. The sensitivity of the PVT-NG sensor to bare 252 Cf is 1.9 counts per second per nanogram (cps/ng) and the sensitivity for 252 Cf surrounded by 2.5 cm of polyethylene is 2.3 cps/ng. The PVT-NG sensor is a proof-of-principal sensor that was not fully optimized. The neutron detector sensitivity could be improved, for instance, by using additional moderator. The PVT-NG detectors and associated electronics are designed to provide improved resolution, gain stability, and performance at high-count rates relative to PVT detectors in typical radiation portals. As well as addressing the needs for neutron detection, these characteristics are also desirable for analysis of the gamma-ray spectra. Accurate isotope identification results were obtained despite the common impression that the absence of photopeaks makes data collected by PVT detectors unsuitable for spectroscopic analysis. The PVT detectors in the PVT-NG unit are used for both gamma-ray and neutron detection, so the sensitive volume exceeds the volume of the detection elements in portals that use dedicated components to detect each type of radiation.

  6. Stereographic images acquired with gamma rays and thermal neutron radiography

    International Nuclear Information System (INIS)

    Souza, Maria Ines Silvani; Almeida, Gevaldo L. de; Furieri, Rosanne C.; Lopes, Ricardo T.

    2011-01-01

    Full text: The inner structure of an object, which should not be submitted to an invasive assay, can only be perceived by using a suitable technique in order to render it transparent. A widely employed technique for this purpose involves the using of a radiation capable to pass through the object, collecting the transmitted radiation by a proper device, which furnishes a radiographic attenuation map of the object. This map, however, does not display the spatial distribution of the inner components of the object, but a convoluted view for each specific attitude of the object with regard to the set beam-detector. A 3D tomographic approach would show that distribution but it would demand a large number of projections requiring special equipment and software, not always available or affordable. In some circumstances however, a 3D tomography can be replaced by a stereographic view of the object under inspection, as done in this work, where instead of tens of radiographic projections, only two of them taken at suitable object attitudes are employed. Once acquired, these projections are properly processed and observed through a red and green eyeglass. For monochromatic images, this methodology requires the transformation of the black and white radiographs into red and white and green and white ones, which are afterwards merged to yield a single image. All the process is carried out with the software Image J . In this work, the Argonauta reactor at the Instituto de Engenharia Nuclear in Rio de Janeiro has been used as a source of thermal neutrons to acquire the neutron radiographic images, as well as to produce 198 Au sources employed in the acquisition of gamma-ray radiographic ones. X-ray or neutron-sensitive imaging plates have been used as detector, which after exposure were developed by a reader using a 0.5μm-diameter laser beam. (author)

  7. Neutron capture prompt gamma-ray activation analysis at the NIST cold neutron research facility

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Zeisler, R; Vincent, D H; Greenberg, R R; Stone, C A; Mackey, E A [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Anderson, D L [Food and Drug Administration, Washington, DC (United States); Clark, D D [Cornell Univ., Ithaca, NY (United States)

    1993-01-01

    An instrument for neutron capture prompt gamma-ray activation analysis (PGAA) has been constructed as part of the Cold Neutron Research Facility at the 20 MW National Institute of Standards and Technology Research Reactor. The neutron fluence rate (thermal equivalent) is 1.5*10[sup 8] n*cm[sup -2]*s[sup -] [sup 1], with negligible fast neutrons and gamma-rays. With compact geometry and hydrogen-free construction, the sensitivity is sevenfold better than an existing thermal instrument. Hydrogen background is thirtyfold lower. (author) 17 refs.; 2 figs.

  8. Integrated neutron/gamma-ray portal monitors for nuclear safeguards

    International Nuclear Information System (INIS)

    Fehlau, P.E.

    1994-01-01

    Radiation monitoring is one nuclear-safeguards measure used to protect against the theft of special nuclear materials (SNM) by pedestrians departing from SNM access areas. The integrated neutron/gamma-ray portal monitor is an ideal radiation monitor for the task when the SNM is plutonium. It achieves high sensitivity for detecting both bare and shielded plutonium by combining two types of radiation detector. One type is a neutron-chamber detector, comprising a large, hollow, neutron moderator that contains a single thermal-neutron proportional counter. The entrance wall of each chamber is thin to admit slow neutrons from plutonium contained in a moderating shield, while the other walls are thick to moderate fast neutrons from bare or lead-shielded plutonium so that they can be detected. The other type of detector is a plastic scintillator that is primarily for detecting gamma rays from small amounts of unshielded plutonium. The two types of detector are easily integrated by making scintillators part of the thick back wall of each neutron chamber or by inserting them into each chamber void. The authors compared the influence of the two methods of integration on detecting neutrons and gamma rays, and they examined the effectiveness of other design factors and the methods for signal detection as well

  9. The measurement of gamma ray induced heating in a mixed neutron and gamma ray environment

    International Nuclear Information System (INIS)

    Chiu, H.K.

    1991-10-01

    The problem of measuring the gamma heating in a mixed DT neutron and gamma ray environment was explored. A new detector technique was developed to make this measurement. Gamma heating measurements were made in a low-Z assembly irradiated with 14-Mev neutrons and (n, n') gammas produced by a Texas Nuclear Model 9400 neutron generator. Heating measurements were made in the mid-line of the lattice using a proportional counter operating in the Continuously-varied Bias-voltage Acquisition mode. The neutron-induced signal was separated from the gamma-induced signal by exploiting the signal rise-time differences inherent to radiations of different linear energy transfer coefficient, which are observable in a proportional counter. The operating limits of this measurement technique were explored by varying the counter position in the low-Z lattice, hence changing the irradiation spectrum observed. The experiment was modelled numerically to help interpret the measured results. The transport of neutrons and gamma rays in the assembly was modelled using the one- dimensional radiation transport code ANISN/PC. The cross-section set used for these calculations was derived from the ENDF/B-V library using the code MC 2 -2 for the case of DT neutrons slowing down in a low-Z material. The calculated neutron and gamma spectra in the slab and the relevant mass-stopping powers were used to construct weighting factors which relate the energy deposition in the counter fill-gas to that in the counter wall and in the surrounding material. The gamma energy deposition at various positions in the lattice is estimated by applying these weighting factors to the measured gamma energy deposition in the counter at those locations

  10. Neutron Assay System for Confinement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

  11. The synchronous active neutron detection system for spent fuel assay

    International Nuclear Information System (INIS)

    Pickrell, M.M.; Kendall, P.K.

    1994-01-01

    The authors have begun to develop a novel technique for active neutron assay of fissile material in spent nuclear fuel. This approach will exploit the unique operating features of a 14-MeV neutron generator developed by Schlumberger. This generator and a novel detection system will be applied to the direct measurement of the fissile material content in spent fuel in place of the indirect measures used at present. The technique they are investigating is termed synchronous active neutron detection (SAND). It closely follows a method that has been used routinely in other branches of physics to detect very small signals in the presence of large backgrounds. Synchronous detection instruments are widely available commercially and are termed open-quotes lock-inclose quotes amplifiers. The authors have implemented a digital lock-in amplifier in conjunction with the Schlumberger neutron generator to explore the possibility of synchronous detection with active neutrons. This approach is possible because the Schlumberger system can operate at up to a 50% duty factor, in effect, a square wave of neutron yield. The results to date are preliminary but quite promising. The system is capable of resolving the fissile material contained in a small fraction of the fuel rods in a cold fuel assembly. It also appears to be quite resilient to background neutron interference. The interrogating neutrons appear to be nonthermal and penetrating. Although a significant amount of work remains to fully explore the relevant physics and optimize the instrument design, the underlying concept appears sound

  12. Imaging of heterogeneous materials by prompt gamma-ray neutron activation analysis

    International Nuclear Information System (INIS)

    Staples, Parrish; Prettyman, Tom; Lestone, John

    1999-01-01

    We have used a Tomographic Gamma Scanner (TGS) to produce tomographic Prompt Gamma-Ray Neutron Activation Imaging of heterogeneous matrices [T.H. Prettyman, R.J. Estep, G.A. Sheppard, Trans. Am. Nucl. Soc. 69 (1993) 183-184]. The TGS was modified by the addition of graphite reflectors that contain isotopic neutron sources for sample interrogation. We are in the process of developing the analysis methodology necessary for a quantitative assay of large containers of heterogeneous material. This nondestructive analysis technique can be used for material characterization and the determination of neutron assay correction factors. The most difficult question to be answered is the determination of the source to sample coupling term. To assist in the determination of the coupling term we have obtained images for a range of samples that are very well characterized; such as, homogenous pseudo one-dimensional samples to three-dimensional heterogeneous samples. We then compare the measurements to Monte Carlo N-particle calculations. For an accurate quantitative measurement it is also necessary to determine the sample gamma-ray self attenuation at higher gamma-ray energies, namely pair production should be incorporated into the analysis codes

  13. Imaging of heterogeneous materials by prompt gamma-ray neutron activation analysis

    International Nuclear Information System (INIS)

    Staples, P.; Prettyman, T.; Lestone, J.

    1998-01-01

    The authors have used a tomographic gamma scanner (TGS) to produce tomographic prompt gamma-ray neutron activation analysis imaging (PGNAA) of heterogeneous matrices. The TGS was modified by the addition of graphite reflectors that contain isotopic neutron sources for sample interrogation. The authors are in the process of developing the analysis methodology necessary for a quantitative assay of large containers of heterogeneous material. This nondestructive analysis (NDA) technique can be used for material characterization and the determination of neutron assay correction factors. The most difficult question to be answered is the determination of the source-to-sample coupling term. To assist in the determination of the coupling term, the authors have obtained images for a range of sample that are very well characterized, such as, homogenous pseudo one-dimensional samples to three-dimensional heterogeneous samples. They then compare the measurements to MCNP calculations. For an accurate quantitative measurement, it is also necessary to determine the sample gamma-ray self attenuation at higher gamma-ray energies, namely pair production should be incorporated into the analysis codes

  14. Computed neutron coincidence counting applied to passive waste assay

    Energy Technology Data Exchange (ETDEWEB)

    Bruggeman, M.; Baeten, P.; De Boeck, W.; Carchon, R. [Nuclear Research Centre, Mol (Belgium)

    1997-11-01

    Neutron coincidence counting applied for the passive assay of fissile material is generally realised with dedicated electronic circuits. This paper presents a software based neutron coincidence counting method with data acquisition via a commercial PC-based Time Interval Analyser (TIA). The TIA is used to measure and record all time intervals between successive pulses in the pulse train up to count-rates of 2 Mpulses/s. Software modules are then used to compute the coincidence count-rates and multiplicity related data. This computed neutron coincidence counting (CNCC) offers full access to all the time information contained in the pulse train. This paper will mainly concentrate on the application and advantages of CNCC for the non-destructive assay of waste. An advanced multiplicity selective Rossi-alpha method is presented and its implementation via CNCC demonstrated. 13 refs., 4 figs., 2 tabs.

  15. Computed neutron coincidence counting applied to passive waste assay

    International Nuclear Information System (INIS)

    Bruggeman, M.; Baeten, P.; De Boeck, W.; Carchon, R.

    1997-01-01

    Neutron coincidence counting applied for the passive assay of fissile material is generally realised with dedicated electronic circuits. This paper presents a software based neutron coincidence counting method with data acquisition via a commercial PC-based Time Interval Analyser (TIA). The TIA is used to measure and record all time intervals between successive pulses in the pulse train up to count-rates of 2 Mpulses/s. Software modules are then used to compute the coincidence count-rates and multiplicity related data. This computed neutron coincidence counting (CNCC) offers full access to all the time information contained in the pulse train. This paper will mainly concentrate on the application and advantages of CNCC for the non-destructive assay of waste. An advanced multiplicity selective Rossi-alpha method is presented and its implementation via CNCC demonstrated. 13 refs., 4 figs., 2 tabs

  16. Neutron and gamma irradiation effects on power semiconductor switches

    International Nuclear Information System (INIS)

    Schwarze, G.E.; Frasca, A.J.

    1990-01-01

    The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN bipolar junction transistors (BJTs), and metal-oxide-semiconductor field effect transistors (MOSFETs)

  17. Gamma ray attenuation coefficient measurement for neutron-absorbent materials

    International Nuclear Information System (INIS)

    Jalali, Majid; Mohammadi, Ali

    2008-01-01

    The compounds Na 2 B 4 O 7 , H 3 BO 3 , CdCl 2 and NaCl and their solutions attenuate gamma rays in addition to neutron absorption. These compounds are widely used in the shielding of neutron sources, reactor control and neutron converters. Mass attenuation coefficients of gamma related to the four compounds aforementioned, in energies 662, 778.9, 867.38, 964.1, 1085.9, 1173, 1212.9, 1299.1,1332 and 1408 keV, have been determined by the γ rays transmission method in a good geometry setup; also, these coefficients were calculated by MCNP code. A comparison between experiments, simulations and Xcom code has shown that the study has potential application for determining the attenuation coefficient of various compound materials. Experiment and computation show that H 3 BO 3 with the lowest average Z has the highest gamma ray attenuation coefficient among the aforementioned compounds

  18. Variation of Neutron Moderating Power on HDPE by Gamma Radiation

    International Nuclear Information System (INIS)

    Park, Kwang June; Ju, June Sik; Kang, Hee Young; Shin, Hee Sung; Kim, Ho Dong

    2009-01-01

    High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a 60 Co source to a level of 10 5 -10 9 rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the 105 rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study

  19. Measuring element for the detection and determination of radiation doses of gamma radiation and neutrons

    International Nuclear Information System (INIS)

    Jahn, W.; Piesch, E.

    1975-01-01

    A measuring element detects and proves both gamma and neutron radiation. The element includes a photoluminescent material which stores gamma radiation and particles of arsenic and phosphorus embedded in the photoluminescent material for detecting neutron radiation. (U.S.)

  20. Neutron and gamma characterization within the FFTF reactor cavity

    International Nuclear Information System (INIS)

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

  1. Evaluation of gamma spectroscopy gauge for uranium-plutonium assay

    International Nuclear Information System (INIS)

    Notea, A.; Segal, Y.

    1975-01-01

    A procedure is presented for the characterization of a gamma passive method for nondestructive analysis of nuclear fuel. The approach provides an organized and systematic way for optimizing the assay system. The key function is the relative resolving power defined as the smallest relative change in the Quantity of radionuclide measured, that may be detected within a certain confidence level. This function is derived for nuclear fuel employing a model based on empirical parameters. The ability to detect changes in fuels of binary and trinary compositions with a 50 cc Ge(Li) at a 1 min counting period is discussed. As an example to a binary composition, an enriched uranium fuel was considered. The 185 keV and 1001 keV gamma lines are used for the assay of 235 U and 238 U respectively. As a trinary composition a plutonium-containing gamma line. The interference of the high energy lines is carefully analyzed, and numerical results are presented. For both cases the range of measurement under specific accuracy demands is determined. The approach described is suitable also for evaluation of other passive as well as active assay methods. (author)

  2. 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.

  3. Statistical precision of delayed-neutron nondestructive assay techniques

    International Nuclear Information System (INIS)

    Bayne, C.K.; McNeany, S.R.

    1979-02-01

    A theoretical analysis of the statistical precision of delayed-neutron nondestructive assay instruments is presented. Such instruments measure the fissile content of nuclear fuel samples by neutron irradiation and delayed-neutron detection. The precision of these techniques is limited by the statistical nature of the nuclear decay process, but the precision can be optimized by proper selection of system operating parameters. Our method is a three-part analysis. We first present differential--difference equations describing the fundamental physics of the measurements. We then derive and present complete analytical solutions to these equations. Final equations governing the expected number and variance of delayed-neutron counts were computer programmed to calculate the relative statistical precision of specific system operating parameters. Our results show that Poisson statistics do not govern the number of counts accumulated in multiple irradiation-count cycles and that, in general, maximum count precision does not correspond with maximum count as first expected. Covariance between the counts of individual cycles must be considered in determining the optimum number of irradiation-count cycles and the optimum irradiation-to-count time ratio. For the assay system in use at ORNL, covariance effects are small, but for systems with short irradiation-to-count transition times, covariance effects force the optimum number of irradiation-count cycles to be half those giving maximum count. We conclude that the equations governing the expected value and variance of delayed-neutron counts have been derived in closed form. These have been computerized and can be used to select optimum operating parameters for delayed-neutron assay devices

  4. Random pulsing of neutron source for inelastic neutron scattering gamma ray spectroscopy

    International Nuclear Information System (INIS)

    Hertzog, R.C.

    1981-01-01

    Method and apparatus are described for use in the detection of inelastic neutron scattering gamma ray spectroscopy. Data acquisition efficiency is enhanced by operating a neutron generator such that a resulting output burst of fast neutrons is maintained for as long as practicably possible until a gamma ray is detected. Upon the detection of a gamma ray the generator burst output is terminated. Pulsing of the generator may be accomplished either by controlling the burst period relative to the burst interval to achieve a constant duty cycle for the operation of the generator or by maintaining the burst period constant and controlling the burst interval such that the resulting mean burst interval corresponds to a burst time interval which reduces contributions to the detected radiation of radiation occasioned by other than the fast neutrons

  5. Prompt gamma cold neutron activation analysis applied to biological materials

    International Nuclear Information System (INIS)

    Rossbach, M.; Hiep, N.T.

    1992-01-01

    Cold neutrons at the external neutron guide laboratory (ELLA) of the KFA Juelich are used to demonstrate their profitable application for multielement characterization of biological materials. The set-up and experimental conditions of the Prompt Gamma Cold Neutron Activation Analysis (PGCNAA) device is described in detail. Results for C, H, N, S, K, B, and Cd using synthetic standards and the 'ratio' technique for calculation are reported for several reference materials and prove the method to be reliable and complementary with respect to the elements being determined by INAA. (orig.)

  6. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Mozin, Vladimir [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Campbell, Luke [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hunt, Alan W. [Idaho State Univ., Pocatello, ID (United States); Reedy, Edward T. [Idaho State Univ., Pocatello, ID (United States); Seipel, Heather A. [Idaho State Univ., Pocatello, ID (United States)

    2015-06-01

    Modeling capabilities were added to an existing framework and codes were adapted as needed for analyzing experiments and assessing application-specific assay concepts including simulation of measurements over many short irradiation/spectroscopy cycles. The code package was benchmarked against the data collected at the IAC for small targets and assembly-scale data collected at LANL. A study of delayed gamma-ray spectroscopy for nuclear safeguards was performed for a variety of assemblies in the extensive NGSI spent fuel library. The modeling results indicate that delayed gamma-ray responses can be collected from spent fuel assemblies with statistical quality sufficient for analyzing their isotopic composition using a 1011 n/s neutron generator and COTS detector instrumentation.

  7. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke; Favalli, Andrea; Hunt, Alan W.; Reedy, Edward T.E.; Seipel, Heather

    2015-01-01

    High-energy, beta-delayed gamma-ray spectroscopy is a potential, non-destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-ray emission was developed and demonstrated on a limited 235 U data set. De-convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an

  8. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Mozin, Vladimir [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Campbell, Luke [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hunt, Alan W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reedy, Edward T.E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Seipel, Heather [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr3 detectors

  9. Gamma-Free Neutron Detector Based upon Lithium Phosphate Nanoparticles

    International Nuclear Information System (INIS)

    Steven Wallace

    2007-01-01

    A gamma-free neutron-sensitive scintillator is needed to enhance radiation sensing and detection for nonproliferation applications. Such a scintillator would allow very large detectors to be placed at the perimeter of spent-fuel storage facilities at commercial nuclear power plants, so that any movement of spontaneously emitted neutrons from spent nuclear fuel or weapons grade plutonium would be noted in real-time. This task is to demonstrate that the technology for manufacturing large panels of fluor-doped plastic containing lithium-6 phosphate nanoparticles can be achieved. In order to detect neutrons, the nanoparticles must be sufficiently small so that the plastic remains transparent. In this way, the triton and alpha particles generated by the capture of the neutron will result in a photon burst that can be coupled to a wavelength shifting fiber (WLS) producing an optical signal of about ten nanoseconds duration signaling the presence of a neutron emitting source

  10. Design of a versatile detector for the detection of charged particles, neutrons and gamma rays. Neutron interaction with the matter

    International Nuclear Information System (INIS)

    Perez P, J.J.

    1991-01-01

    The Fostron detector detects charged particles, neutrons and gamma rays with a reasonable discrimination power. Because the typical detectors for neutrons present a great uncertainty in the detection, this work was focused mainly to the neutron detection in presence of gamma radiation. Also there are mentioned the advantages and disadvantages of the Fostron detector

  11. Gamma and Neutron Irradiation of Semitransparent Amorphous Silicon Sensors

    International Nuclear Information System (INIS)

    Carabe, J.; Fernandez, M. G.; Ferrando, A.; Fuentes, J.; Gandia, J.; Josa, M. I.; Molinero, A.; Oller, J. C.; Arce, P.; Calvo, E.; Figueroa, C. F.; Garcia, N.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A. L.; Fenyvesi, A.; Molnar, J.; Sohler, D.

    1999-12-01

    Semitransparent amorphous silicon sensors are key elements for laser light 2D position reconstruction in the CMS multipoint alignment link system. Some of the sensors have to work in very hard radiation environment. We have irradiated with gammas, up to 10 Mrad, and neutrons, up to 10 ''14 cm''-2, two different type of sensors and measured their change in performance. (Author) 10 refs

  12. Results on Neutron and Gamma Irradiation of Electrolytic Tilmeters

    International Nuclear Information System (INIS)

    Calderon, A.; Calvo, E.; Figueroa, C. F.; Martinez-Rivero, C.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A. L.; Alberdi, J.; Arce, P.; Barcala, J. M.; Fernando, A.; Fuentes, J.; Josa, M. I.; Luque, J. M.; Molinero, A.; Navarrate, J.; Valdivieso, P.; Fenyvesi, A.; Molnar, J.

    2004-01-01

    We report on irradiation studies done to a sample of high precision electrolytic tiltmeters with gamma-rays, up to a maximum dose of 150 kGy, an neutrons, up to a maximum fluence 1.5x10''14 cm''2. The effect of the irradiation on their performance is discussed. (Author) 19 refs

  13. Results on Neutron and Gamma Irradiation of Electrolytic Tilmeters

    Energy Technology Data Exchange (ETDEWEB)

    Calderon, A.; Calvo, E.; Figueroa, C. F.; Martinez-Rivero, C.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A. L.; Alberdi, J.; Arce, P.; Barcala, J. M.; Fernando, A.; Fuentes, J.; Josa, M. I.; Luque, J. M.; Molinero, A.; Navarrate, J.; Valdivieso, P.; Fenyvesi, A.; Molnar, J.

    2004-07-01

    We report on irradiation studies done to a sample of high precision electrolytic tiltmeters with gamma-rays, up to a maximum dose of 150 kGy, an neutrons, up to a maximum fluence 1.5x10''14 cm''2. The effect of the irradiation on their performance is discussed. (Author) 19 refs.

  14. Gamma rays from fast neutron capture in silicon and sulphur

    International Nuclear Information System (INIS)

    Lindholm, A.; Nilsson, L.; Bergqvist, I.

    1975-01-01

    Gamma-ray spectra from neutron capture in natural samples of silicon and sulphur have been recorded at eight neutron energies between 4 and 15 MeV. Time-of-flight techniques were used to improve the signal-to-background ratio and the gamma radiation was detected by a large NaI(Tl) scintillator. Cross sections have been determined for transitions to individual (or groups of) levels in the final nucleus. Calculations based on the direct-semidirect model show that this model gives a reasonable description of the shapes of the gamma-ray spectra, but fails to account for observed excitation functions. The inclusion of the compound-nucleus capture process gives a conclusive improvement in the description of the excitation functions, in particular at low neutron energies. The ability of the compound-nucleus model to account for the shapes of the gamma-ray spectra is as good as that of the direct-semidirect model. At higher neutron energies, an improvement is obtained for transitions to the region of weakly bound levels, where the single-particle structure is poorly known. (Auth.)

  15. Semiconductor dosimetry system for gamma and neutron radiation

    International Nuclear Information System (INIS)

    Savic, Z.; Pavlovic, Z.

    1995-01-01

    The semiconductor dosimetry system for gamma and neutron radiation based on pMOS transistor and PIN diode is described. It is intended for tactical or accidental personal dosimetry. The production steps are given. The temperature, dose and time (fading) response are reported. Hardware and software requirements which are needed for obtaining the desired measurement error are pointed. (author)

  16. Deduction of solar neutron fluences from large gamma-ray flares

    International Nuclear Information System (INIS)

    Yoshimori, Masato; Watanabe, Hiroyuki; Takahashi, Kazuyoshi.

    1986-01-01

    Solar neutron fluences from large gamma-ray flares are deduced from accelerated proton spectra and numbers derived from the gamma-ray observations. The deduced solar neutron fluences range from 1 to 200 neutrons cm -2 . The present result indicates a possibility that high sensitivity ground-based neutron monitors can detect solar neutron events, just as detected by the Jungfraujoch and Rome neutron monitors. (author)

  17. Two-dimensional neutron scintillation detector with optimal gamma discrimination

    International Nuclear Information System (INIS)

    Kanyo, M.; Reinartz, R.; Schelten, J.; Mueller, K.D.

    1993-01-01

    The gamma sensitivity of a two-dimensional scintillation neutron detector based on position sensitive photomultipliers (Hamamatsu R2387 PM) has been minimized by a digital differential discrimination unit. Since the photomultiplier gain is position-dependent by ±25% a discrimination unit was developed where digital upper and lower discrimination levels are set due to the position-dependent photomultiplier gain obtained from calibration measurements. By this method narrow discriminator windows can be used to reduce the gamma background drastically without effecting the neutron sensitivity of the detector. The new discrimination method and its performance tested by neutron measurements will be described. Experimental results concerning spatial resolution and γ-sensitivity are presented

  18. Gamma-Ray Bursts from Neutron Star Kicks

    Science.gov (United States)

    Huang, Y. F.; Dai, Z. G.; Lu, T.; Cheng, K. S.; Wu, X. F.

    2003-09-01

    The idea that gamma-ray bursts might be a phenomenon associated with neutron star kicks was first proposed by Dar & Plaga. Here we study this mechanism in more detail and point out that the neutron star should be a high-speed one (with proper motion larger than ~1000 km s-1). It is shown that the model agrees well with observations in many aspects, such as the energetics, the event rate, the collimation, the bimodal distribution of durations, the narrowly clustered intrinsic energy, and the association of gamma-ray bursts with supernovae and star-forming regions. We also discuss the implications of this model on the neutron star kick mechanism and suggest that the high kick speed was probably acquired as the result of the electromagnetic rocket effect of a millisecond magnetar with an off-centered magnetic dipole.

  19. Polycrystalline Materials as a Cold Neutron and Gamma Radiation Filter

    International Nuclear Information System (INIS)

    Habib, N.

    2009-01-01

    The total neutron cross-section of polycrystalline beryllium, graphite and iron has been calculated beyond their cut-off wavelength using a general formula. The computer Cold Filter code was developed in order to provide the required calculations. The code also permits the calculation of attenuation of reactor gamma radiation, The calculated neutron transmissions through polycrystalline Be graphite and iron at different temperatures were compared with the experimental data measured at the ETRR-1 reactor using two TOF spectrometers. An overall agreement is obtained between the formula fits and experimental data at different temperatures. A feasibility study is carried on using polycrystalline Be, graphite and iron an efficient filter for cold neutrons and gamma radiation.

  20. Test plan for a live drum survey using the gamma-neutron sensor

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Roybal, L.G.; Thompson, D.N.

    1995-07-01

    This plan describes performance tests to be made with the Gamma/Neutron Sensor (GNS), which that was designed and built for infield assay at an excavation site. The performance tests will be performed in Building WMF-628 in the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory on stored 55-gal drums of transuranic waste from the Rocky Flats Plant. The GNS is mounted on a wooden pallet that will allow horizontal and vertical scans of the stacked drums. Scanning speed and GNS sensitivity for gamma and neutron radiation fields will be estimated. Effects of temperature, electronic, and acoustic noise will be evaluated. Two- and three-dimensional plots of radiation field as a function of position will be developed from the data

  1. Prototype fast neutron counter for the assay of impure plutonium

    International Nuclear Information System (INIS)

    Wachter, J.R.; Adams, E.L.; Ensslin, N.

    1987-01-01

    A fast coincident neutron counter using liquid scintillators and gamma-ray/neutron pulse-shape discrimination has been constructed for the analysis of plutonium samples with unknown self-multiplication and (α,n) production. The counter was used to measure plutonium-bearing materials that cover a range of masses and (α,n) reaction rates of importance to the safeguards community. Measured values of the 240 Pu effective mass differed, on average, from their declared values by 0.4% for plutonium oxides and by -2.2% for metal and MgO-loaded samples. Poorer results were obtained for materials with large (α,n) reaction rates and low self-multiplication such as plutonium ash and plutonium fluoride

  2. Use of digital computers for correction of gamma method and neutron-gamma method indications

    International Nuclear Information System (INIS)

    Lakhnyuk, V.M.

    1978-01-01

    The program for the NAIRI-S computer is described which is intended for accounting and elimination of the effect of by-processes when interpreting gamma and neutron-gamma logging indications. By means of slight corrections it is possible to use the program as a mathematical basis for logging diagram standardization by the method of multidimensional regressive analysis and estimation of rock reservoir properties

  3. Application of the decoupling scheme on complex neutron-gamma shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

    2000-03-01

    Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

  4. Inhomogeneity of neutron and gamma-ray attenuation in biological shields

    Energy Technology Data Exchange (ETDEWEB)

    El-bakkoush, F A; El-Ghobary, A M; Megahid, R M [Reactor and Neutron physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)

    1997-12-31

    Measurements have been carried-out to investigate the attenuation properties of some materials which are used as biological shields around nuclear radiation sources. Investigation was performed by measuring the transmitted fast neutron and gamma-spectra through cylindrical samples of magnetite- limonite, steel and cellulose shields. The neutron and gamma spectra were measured by a neutron-gamma spectrometer with stilbene scintillator. Discrimination between neutron and gamma pulses was achieved by a discrimination method. The obtained results are displayed in the form of neutron and gamma spectra and attenuation relations which are used to derive the total macroscopic cross-sections for neutrons and total linear attenuation coefficients for gamma-rays. The values of neutron and gamma relaxation lengths are also derived for the investigated materials. 10 figs., 1 tabs.

  5. Prompt gamma-based neutron dosimetry for Am-Be and other workplace neutron spectra

    International Nuclear Information System (INIS)

    Udupi, Ashwini; Panikkath, Priyada; Sarkar, P.K.

    2016-01-01

    A new field-deployable technique for estimating the neutron ambient dose equivalent H*(10) by using the measured prompt gamma intensities emitted from borated high-density polyethylene (BHDPE) and the combination of normal HDPE and BHDPE with different configurations have been evaluated in this work. Monte Carlo simulations using the FLUKA code has been employed to calculate the responses from the prompt gammas emitted due to the monoenergetic neutrons interacting with boron, hydrogen, and carbon nuclei. A suitable linear combination of these prompt gamma responses (dose conversion coefficient (DCC)-estimated) is generated to approximate the International Commission on Radiological Protection provided DCC using the cross-entropy minimization technique. In addition, the shape and configurations of the HDPE and BHDPE combined system are optimized using the FLUKA code simulation results. The proposed method is validated experimentally, as well as theoretically, using different workplace neutron spectra with a satisfactory outcome. (author)

  6. System and plastic scintillator for discrimination of thermal neutron, fast neutron, and gamma radiation

    Science.gov (United States)

    Zaitseva, Natalia P.; Carman, M. Leslie; Faust, Michelle A.; Glenn, Andrew M.; Martinez, H. Paul; Pawelczak, Iwona A.; Payne, Stephen A.

    2017-05-16

    A scintillator material according to one embodiment includes a polymer matrix; a primary dye in the polymer matrix, the primary dye being a fluorescent dye, the primary dye being present in an amount of 3 wt % or more; and at least one component in the polymer matrix, the component being selected from a group consisting of B, Li, Gd, a B-containing compound, a Li-containing compound and a Gd-containing compound, wherein the scintillator material exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons and gamma rays. A system according to one embodiment includes a scintillator material as disclosed herein and a photodetector for detecting the response of the material to fast neutron, thermal neutron and gamma ray irradiation.

  7. The Neutron-Gamma Pulse Shape Discrimination Method for Neutron Flux Detection in the ITER

    International Nuclear Information System (INIS)

    Xu Xiufeng; Li Shiping; Cao Hongrui; Yin Zejie; Yuan Guoliang; Yang Qingwei

    2013-01-01

    The neutron flux monitor (NFM), as a significant diagnostic system in the International Thermonuclear Experimental Reactor (ITER), will play an important role in the readings of a series of key parameters in the fusion reaction process. As the core of the main electronic system of the NFM, the neutron-gamma pulse shape discrimination (n-γ PSD) can distinguish the neutron pulse from the gamma pulse and other disturbing pulses according to the thresholds of the rising time and the amplitude pre-installed on the board, the double timing point CFD method is used to get the rising time of the pulse. The n-γ PSD can provide an accurate neutron count. (magnetically confined plasma)

  8. Gamma-ray measurements at the WNR white neutron source

    International Nuclear Information System (INIS)

    Nelson, R.O.; Wender, S.A.; Mayo, D.R.

    1994-01-01

    Photon production data have been acquired in the incident neutron energy range, 1 n γ 56 Fe, and 207,208 Pb. These data are useful both for testing nuclear reaction models at intermediate energies and for numerous applied purposes. BGO detectors do not have the good energy resolution of Ge detectors, but have much greater detection efficiency for gamma rays with energies greater than a few MeV. We have used an array of 5 BGO detectors to measure cross sections and angular distributions for photon production from C and N. A large, well-shielded BGO detector has been used to measure fast neutron capture in the giant resonance region with a maximum gamma-ray energy of 52 MeV. We present results of our study of the isovector giant quadrupole resonance in 41 Ca via these capture measurements. Recent measurements of inclusive photon spectra from our neutron proton Bremsstrahlung experiment have been made using a gamma-ray telescope to detect gamma-rays in the energy range, 40 γ < 300 MeV. This detector is briefly described. The advantages and disadvantages of these detector systems are discussed using examples from our measurements. The status of current measurements is presented

  9. Gamma ray attenuation coefficient measurement for neutron-absorbent materials

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, Majid [Isfahan Nuclear Science and Technology Research Institute (NSTRT), Reactor and Accelerators Research and Development School, Atomic Energy Organization (Iran, Islamic Republic of)], E-mail: m_jalali@entc.org.ir; Mohammadi, Ali [Faculty of Science, Department of Physics, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2008-05-15

    The compounds Na{sub 2}B{sub 4}O{sub 7}, H{sub 3}BO{sub 3}, CdCl{sub 2} and NaCl and their solutions attenuate gamma rays in addition to neutron absorption. These compounds are widely used in the shielding of neutron sources, reactor control and neutron converters. Mass attenuation coefficients of gamma related to the four compounds aforementioned, in energies 662, 778.9, 867.38, 964.1, 1085.9, 1173, 1212.9, 1299.1,1332 and 1408 keV, have been determined by the {gamma} rays transmission method in a good geometry setup; also, these coefficients were calculated by MCNP code. A comparison between experiments, simulations and Xcom code has shown that the study has potential application for determining the attenuation coefficient of various compound materials. Experiment and computation show that H{sub 3}BO{sub 3} with the lowest average Z has the highest gamma ray attenuation coefficient among the aforementioned compounds.

  10. MCNP efficiency calculations of INEEL passive active neutron assay system for simulated TRU waste assays

    International Nuclear Information System (INIS)

    Yoon, W.Y.; Meachum, T.R.; Blackwood, L.G.; Harker, Y.D.

    2000-01-01

    The Idaho National Engineering and Environmental Laboratory Stored Waste Examination Pilot Plant (SWEPP) passive active neutron (PAN) radioassay system is used to certify transuranic (TRU) waste drums in terms of quantifying plutonium and other TRU element activities. Depending on the waste form involved, significant systematic and random errors need quantification in addition to the counting statistics. To determine the total uncertainty of the radioassay results, a statistical sampling and verification approach has been developed. In this approach, the total performance of the PAN nondestructive assay system is simulated using the computer models of the assay system, and the resultant output is compared with the known input to assess the total uncertainty. The supporting steps in performing the uncertainty analysis for the passive assay measurements in particular are as follows: (1) Create simulated waste drums and associated conditions; (2) Simulate measurements to determine the basic counting data that would be produced by the PAN assay system under the conditions specified; and (3) Apply the PAN assay system analysis algorithm to the set of counting data produced by simulating measurements to determine the measured plutonium mass. The validity of this simulation approach was verified by comparing simulated output against results from actual measurements using known plutonium sources and surrogate waste drums. The computer simulation of the PAN system performance uses the Monte Carlo N-Particle (MCNP) Code System to produce a neutron transport calculation for a simulated waste drum. Specifically, the passive system uses the neutron coincidence counting technique, utilizing the spontaneous fission of 240 Pu. MCNP application to the SWEPP PAN assay system uncertainty analysis has been very useful for a variety of waste types contained in 208-ell drums measured by a passive radioassay system. The application of MCNP to the active radioassay system is also feasible

  11. Improved neutron-gamma discrimination for a 3He neutron detector using subspace learning methods

    Science.gov (United States)

    Wang, C. L.; Funk, L. L.; Riedel, R. A.; Berry, K. D.

    2017-05-01

    3He gas based neutron Linear-Position-Sensitive Detectors (LPSDs) have been used for many neutron scattering instruments. Traditional Pulse-height Analysis (PHA) for Neutron-Gamma Discrimination (NGD) resulted in the neutron-gamma efficiency ratio (NGD ratio) on the order of 105-106. The NGD ratios of 3He detectors need to be improved for even better scientific results from neutron scattering. Digital Signal Processing (DSP) analyses of waveforms were proposed for obtaining better NGD ratios, based on features extracted from rise-time, pulse amplitude, charge integration, a simplified Wiener filter, and the cross-correlation between individual and template waveforms of neutron and gamma events. Fisher Linear Discriminant Analysis (FLDA) and three Multivariate Analyses (MVAs) of the features were performed. The NGD ratios are improved by about 102-103 times compared with the traditional PHA method. Our results indicate the NGD capabilities of 3He tube detectors can be significantly improved with subspace-learning based methods, which may result in a reduced data-collection time and better data quality for further data reduction.

  12. Optimum filter-based discrimination of neutrons and gamma rays

    International Nuclear Information System (INIS)

    Amiri, Moslem; Prenosil, Vaclav; Cvachovec, Frantisek

    2015-01-01

    An optimum filter-based method for discrimination of neutrons and gamma-rays in a mixed radiation field is presented. The existing filter-based implementations of discriminators require sample pulse responses in advance of the experiment run to build the filter coefficients, which makes them less practical. Our novel technique creates the coefficients during the experiment and improves their quality gradually. Applied to several sets of mixed neutron and photon signals obtained through different digitizers using stilbene scintillator, this approach is analyzed and its discrimination quality is measured. (authors)

  13. Measurements and analysis of neutron and gamma noise in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van; Kleiss, E.B.J.

    1985-01-01

    Neutron and gamma sensitive collectrons (self-powered detectors) have been designed for incore noise measurements in BWRs. A so-called twin-type has been developed for measurements of two-phase flow characteristics and detailed axial velocity distributions. Construction aspects of the twin detectors are discussed. An analysis is presented of the response of both detector types to incore parametric fluctuations. This analysis is based on detector response functions which provide an insight into the 'field of view' of the two types. The results are supported by experimental verifications; it is shown that incore gamma detectors provide useful additional information about two-phase flow in a BWR. (author)

  14. Development of advanced sensing system for antipersonnel mines with neutron capture gamma-ray analysis

    International Nuclear Information System (INIS)

    Iguchi, Tetsuo

    2006-01-01

    Neutron induced prompt gamma-ray analysis (NPGA) for survey of antipersonnel landmines is developed. A concept of sensor system with compact strong accelerator neutron source, simulation of detection and simulation results by trial examinations are stated. The measurement principles, objects, system construction, development of compact accelerator neutron source and high performance neutron capture gamma-ray detector, simulation of detection of landmine are reported. It can detect 10.8 MeV gamma-rays and estimate the incident angle of gamma-ray. Schematic layouts of the compact accelerator neutron resource, the compact Compton gamma camera and sensor unit, the estimation principle of incident angle of gamma-ray, experiments and comparison between the experimental results and the estimation results, a preliminary trial experiment system for sensing antipersonnel mines with neutron capture gamma-ray analysis are illustrated. (S.Y.)

  15. Neutron and Gamma Imaging for National Security Applications

    Science.gov (United States)

    Hornback, Donald

    2017-09-01

    The Department of Energy, National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D/NA-22) possesses, in part, the mission to develop technologies in support of nuclear security efforts in coordination with other U.S. government entities, such as the Department of Defense and the Department of Homeland Security. DNN R&D has long supported research in nuclear detection at national labs, universities, and through the small business innovation research (SBIR) program. Research topics supported include advanced detector materials and electronics, detection algorithm development, and advanced gamma/neutron detection systems. Neutron and gamma imaging, defined as the directional detection of radiation as opposed to radiography, provides advanced detection capabilities for the NNSA mission in areas of emergency response, international safeguards, and nuclear arms control treaty monitoring and verification. A technical and programmatic overview of efforts in this field of research will be summarized.

  16. Effect of neutron and gamma irradiation on magnetic bubble memories

    International Nuclear Information System (INIS)

    Cambou, B.

    1981-06-01

    Many years of research preceeded the introduction of magnetic bubble memories (M.B.M.) into the memory components market. They are used as bulk storage memories principally for their non volatile characteristics under irradiation. A physical and technological description of MBM is given in the first part of the text together with the results of work on their vulnerability when subjected to irradiation. Permanent damage caused by neutrons and gamma radiation on thin magnetic layers is then studied. A theoretical analysis on the stability of bubbles based on the results of pulsed laser experiments is given. The stability of the information stored in a commercially available MBM subjected to neutron and gamma irradiation (MBM - TIB 203 of 92 kBits, Texas) is described in the last part of the text. The vulnerability thresholds determined for the MBM are too high for them to be used in a radioactive environment with an improved electronic control system [fr

  17. Simultaneous analysis of qualitative parameters of solid fuel using complex neutron gamma method

    International Nuclear Information System (INIS)

    Dombrovskij, V.P.; Ajtsev, N.I.; Ryashchikov, V.I.; Frolov, V.K.

    1983-01-01

    A study was made on complex neutron gamma method for simultaneous analysis of carbon content, ash content and humidity of solid fuel according to gamma radiation of inelastic fast neutron scattering and radiation capture of thermal neutrons. Metrological characteristics of pulse and stationary neutron gamma methods for determination of qualitative solid fuel parameters were analyzed, taking coke breeze as an example. Optimal energy ranges of gamma radiation detection (2-8 MeV) were determined. The advantages of using pulse neutron generator for complex analysis of qualitative parameters of solid fuel in large masses were shown

  18. Bulk moisture determination in building materials by fast neutron/gamma technique

    International Nuclear Information System (INIS)

    Padron Diaz, I.; Felipe Desdin, L.; Martin Hernandez, G.; Shtejer, K.; Perez Tamayo, N.; Ceballos, C.; Lemus, O.

    1998-01-01

    Fast Neutron/Gamma Transmission technique has been improved to allow to measure moisture content in building materials. In order to improve fast neutron/gamma discrimination in the transmission system employing the NE-213 scintillation detector a pulse shape discrimination system was constructed at the CEADEN. A separate neutron/gamma detection approach was used with neutron transmission measurement using an Am-Be neutron source and a BF 3 detector and gamma transmission measurement using a collimated 137 Cs source and a NaI scintillator

  19. Synergistic effects of neutron and gamma ray irradiation of a commercial CHMOS microcontroller

    International Nuclear Information System (INIS)

    Xiao-Ming, Jin; Ru-Yu, Fan; Wei, Chen; Dong-Sheng, Lin; Shan-Chao, Yang; Xiao-Yan, Bai; Yan, Liu; Xiao-Qiang, Guo; Gui-Zhen, Wang

    2010-01-01

    This paper presents the experimental results of a combined irradiation environment of neutron and gamma rays on 80C196KC20, which is a 16-bit high performance member of the MCS96 microcontroller family. The electrical and functional tests were made in three irradiation environments: neutron, gamma rays, combined irradiation of neutron and gamma rays. The experimental results show that the neutron irradiation can affect the total ionizing dose behaviour. Compared with the single radiation environment, the microcontroller exhibits considerably more severe degradation in neutron and gamma ray synergistic irradiation. This phenomenon may cause a significant hardness assurance problem. (condensed matter: structure, thermal and mechanical properties)

  20. Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields

    International Nuclear Information System (INIS)

    Fujibuchi, T.; Tanabe, Y.; Sakae, T.; Terunuma, T.; Isobe, T.; Kawamura, H.; Yasuoka, K.; Matsumoto, T.; Harano, H.; Nishiyama, J.; Masuda, A.; Nohtomi, A.

    2011-01-01

    In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field. (authors)

  1. Resonant production of $\\gamma$ rays in jolted cold neutron stars

    CERN Document Server

    Kusenko, A

    1998-01-01

    Acoustic shock waves passing through colliding cold neutron stars can cause repetitive superconducting phase transitions in which the proton condensate relaxes to its equilibrium value via coherent oscillations. As a result, a resonant non-thermal production of gamma rays in the MeV energy range with power up to 10^(52) erg/s can take place during the short period of time before the nuclear matter is heated by the shock waves.

  2. Development of a lion-specific interferon-gamma assay.

    Science.gov (United States)

    Maas, M; van Kooten, P J S; Schreuder, J; Morar, D; Tijhaar, E; Michel, A L; Rutten, V P M G

    2012-10-15

    The ongoing spread of bovine tuberculosis (BTB) in African free-ranging lion populations, for example in the Kruger National Park, raises the need for diagnostic assays for BTB in lions. These, in addition, would be highly relevant for zoological gardens worldwide that want to determine the BTB status of their lions, e.g. for translocations. The present study concerns the development of a lion-specific IFN-γ assay, following the production and characterization of monoclonal antibodies specific for lion interferon-gamma (IFN-γ). Recombinant lion IFN-γ (rLIFN-γ) was produced in mammalian cells and used to immunize mice to establish hybridoma cell lines producing monoclonal antibodies. These were used to develop a sensitive, lion IFN-γ-specific capture ELISA, able to detect rLIFN-γ to the level of 160 pg/ml. Recognition of native lion IFN-γ was shown in an initial assessment of supernatants of mitogen stimulated whole blood cultures of 11 known BTB-negative lions. In conclusion, the capture ELISA shows potential as a diagnostic assay for bovine tuberculosis in lions. Preliminary results also indicate the possible use of the test for other (feline) species. Copyright © 2012 Elsevier B.V. All rights reserved.

  3. Design and operation of a passive neutron monitor for assaying the TRU content of solid wastes

    International Nuclear Information System (INIS)

    Brodzinski, R.L.; Brown, D.P.; Rieck, H.G. Jr.; Rogers, L.A.

    1984-02-01

    A passive neutron monitor has been designed and built for determining the residual transuranic (TRU) and plutonium content of chopped leached fuel hulls and other solid wastes from spent Fast Flux Test Facility (FFTF) fuel. The system was designed to measure as little as 8 g of plutonium or 88 mg of TRU in a waste package as large as a 208-l drum which could be emitting up to 220,000 R/hr of gamma radiation. For practical purposes, maximum assay times were chosen to be 10,000 sec. The monitor consists of 96 10 BF 3 neutron sensitive proportional counting tubes each 5.08 cm in diameter and 183 cm in active length. Tables of neutron emission rates from both spontaneous fission and (α,n) reactions on oxygen are given for all contributing isotopes expected to be present in spent FFTF fuel. Tables of neutron yeilds from isotopic compositions predicted for various exposures and cooling times are also given. Methods of data reduction and sources, magnitude, and control of errors are discussed. Backgrounds and efficiencies have been measured and are reported. A section describing step-by-step operational procedures is included. Guidelines and procedures for quality control and troubleshooting are also given. 13 references, 15 figures, 4 tables

  4. Reactivation of neutron killed mammalian cells by gamma irradiation: The observations, possible mechanism and implication

    International Nuclear Information System (INIS)

    Calkins, J.; Harrison, W.; Einspenner, M.

    1990-01-01

    We have observed that combinations of neutron plus gamma ray exposure can significantly increase the colony forming ability of monkey and human cell cultures over the neutron dose alone. The 'reactivation' of neutron killed mammalian cells by gamma rays is analogous to observations made in lower eukaryotic organisms and fits the pattern termed 'T repair' previously postulated for yeast and protozoans. (orig.)

  5. Sparse image representation for jet neutron and gamma tomography

    Energy Technology Data Exchange (ETDEWEB)

    Craciunescu, T. [EURATOM-MEdC Association, Institute for Laser, Plasma and Radiation Physics, Bucharest (Romania); Kiptily, V. [EURATOM/CCFE Association, Culham Science Centre, Abingdon (United Kingdom); Murari, A. [Consorzio RFX, Associazione EURATOM-ENEA per la Fusione, Padova (Italy); Tiseanu, I.; Zoita, V. [EURATOM-MEdC Association, Institute for Laser, Plasma and Radiation Physics, Bucharest (Romania)

    2013-10-15

    Highlights: •A new tomographic method for the reconstruction of the 2-D neutron and gamma emissivity on JET. •The method is based on the sparse representation of the reconstructed image in an over-complete dictionary. •Several techniques, based on a priori information are used to regularize this highly limited data set tomographic problem. •The proposed method provides good reconstructions in terms of shapes and resolution. -- Abstract: The JET gamma/neutron profile monitor plasma coverage of the emissive region enables tomographic reconstruction. However, due to the availability of only two projection angles and to the coarse sampling, tomography is a highly limited data set problem. A new reconstruction method, based on the sparse representation of the reconstructed image in an over-complete dictionary, has been developed and applied to JET neutron/gamma tomography. The method has been tested on JET experimental data and significant results are presented. The proposed method provides good reconstructions in terms of shapes and resolution.

  6. Application of neutron-gamma analysis for determination of C/N ratio in compost

    Science.gov (United States)

    Neutron-gamma analysis is based on the acquisition of gamma rays from neutron irradiated study objects. The intensity and energy of the registered gamma rays gives information on the types and amounts of elements in the studied object. The use of this method for measurements of soil carbon demonstra...

  7. Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination

    Science.gov (United States)

    2016-06-01

    Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination Distribution Statement A. Approved for public release; distribution is...Final Technical Report BRBAA08-Per5-Y-1-2-0030 Title: “Gadolinium-Based GaN for Neutron Detection with Gamma Discrimination ” Grant...Analysis  .............................................................................................  23   6.   Gamma-ray Discrimination

  8. Assessment of the neutron component in a neutron-gamma field of a californium-252 source

    International Nuclear Information System (INIS)

    Tetteh, G.K.

    1978-12-01

    Experiments have been performed to determine the percentages of the different components in the radiation field of californium-252 which has now some clinical applications. Using Rossi Chambers in conjunction with absorption investigations involving lead and aluminium thimbles, it is observed that the dose rates due to the different components are: neutrons 54%; gammas 30%; betas 16%

  9. EPR dosimetry in a mixed neutron and gamma radiation field.

    Science.gov (United States)

    Trompier, F; Fattibene, P; Tikunov, D; Bartolotta, A; Carosi, A; Doca, M C

    2004-01-01

    Suitability of Electron Paramagnetic Resonance (EPR) spectroscopy for criticality dosimetry was evaluated for tooth enamel, mannose and alanine pellets during the 'international intercomparison of criticality dosimetry techniques' at the SILENE reactor held in Valduc in June 2002, France. These three materials were irradiated in neutron and gamma-ray fields of various relative intensities and spectral distributions in order to evaluate their neutron sensitivity. The neutron response was found to be around 10% for tooth enamel, 45% for mannose and between 40 and 90% for alanine pellets according their type. According to the IAEA recommendations on the early estimate of criticality accident absorbed dose, analyzed results show the EPR potentiality and complementarity with regular criticality techniques.

  10. Performance test results of noninvasive characterization of RCRA surrogate waste by prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Propp, W.A.

    1997-11-01

    A performance evaluation to determine the feasibility of using prompt gamma neutron activation analysis (PGNAA) for noninvasive, quantitative assay of mixed waste containers was sponsored by DOE's Office of Technology Development (OTD), the Mixed Waste Focus Area (MWFA), and the Idaho National Engineering and Environmental Laboratory (INEEL). The evaluation was conducted using a surrogate waste, based on Portland cement, that was spiked with three RCRA metals, mercury, cadmium, and lead. The results indicate that PGNAA has potential as a process monitor. However, further development is required to improve its sensitivity to meet regulatory requirements for determination of these RCRA metals

  11. Calibration of the delayed-gamma neutron activation facility

    International Nuclear Information System (INIS)

    Ma, R.; Zhao, X.; Rarback, H.M.; Yasumura, S.; Dilmanian, F.A.; Moore, R.I.; Lo Monte, A.F.; Vodopia, K.A.; Liu, H.B.; Economos, C.D.; Nelson, M.E.; Aloia, J.F.; Vaswani, A.N.; Weber, D.A.; Pierson, R.N. Jr.; Joel, D.D.

    1996-01-01

    The delayed-gamma neutron activation facility at Brookhaven National Laboratory was originally calibrated using an anthropomorphic hollow phantom filled with solutions containing predetermined amounts of Ca. However, 99% of the total Ca in the human body is not homogeneously distributed but contained within the skeleton. Recently, an artificial skeleton was designed, constructed, and placed in a bottle phantom to better represent the Ca distribution in the human body. Neutron activation measurements of an anthropomorphic and a bottle (with no skeleton) phantom demonstrate that the difference in size and shape between the two phantoms changes the total body calcium results by less than 1%. To test the artificial skeleton, two small polyethylene jerry-can phantoms were made, one with a femur from a cadaver and one with an artificial bone in exactly the same geometry. The femur was ashed following the neutron activation measurements for chemical analysis of Ca. Results indicate that the artificial bone closely simulates the real bone in neutron activation analysis and provides accurate calibration for Ca measurements. Therefore, the calibration of the delayed-gamma neutron activation system is now based on the new bottle phantom containing an artificial skeleton. This change has improved the accuracy of measurement for total body calcium. Also, the simple geometry of this phantom and the artificial skeleton allows us to simulate the neutron activation process using a Monte Carlo code, which enables us to calibrate the system for human subjects larger and smaller than the phantoms used as standards. copyright 1996 American Association of Physicists in Medicine

  12. Neutron detector with gamma compensated cable

    International Nuclear Information System (INIS)

    Warren, H.D.

    1975-01-01

    An illustrative embodiment of the invention describes a technique for essentially eliminating the radiation induced background currents that are generated in the cable that connects an ''in-core'' neutron detector to an electrical terminal that is outside of the reactor's radiation field. This undesirable radiation-induced cable current is suppressed through an appropriate selection of conductor and cable sheath materials and sizes that generally satisfy the equation: Z/sub l/sup n/d/sub l/ = Z/sub s/sup m/d/sub s/ where Z is the atomic number of the material; d is a characteristic of the size of the cable component; m and/n have values between 1 and 5 to express the electron emissivity of the cable component from photoelectric and Compton effects; l represents the conductor; and s represents the sheath. Thus, the radiation-generated electrons emitted from the conductor and the oppositely-directed electrons emitted from the inner surface of the cable sheath are mutually cancelled if this equation is satisfied. A typical cable that does meet this criterion at low temperatures has a centrally disposed Zircaloy-2 inner conductor of 0.011 inch diameter, an annular insulation of magnesium oxide powder compacted to 100 percent density, and an Inconel sheath with an outside diameter of 0.062 inch and 0.011 inch wall thickness. (auth)

  13. Designing a new type of neutron detector for neutron and gamma-ray discrimination via GEANT4

    International Nuclear Information System (INIS)

    Shan, Qing; Chu, Shengnan; Ling, Yongsheng; Cai, Pingkun; Jia, Wenbao

    2016-01-01

    Design of a new type of neutron detector, consisting of a fast neutron converter, plastic scintillator, and Cherenkov detector, to discriminate 14-MeV fast neutrons and gamma rays in a pulsed n–γ mixed field and monitor their neutron fluxes is reported in this study. Both neutrons and gamma rays can produce fluorescence in the scintillator when they are incident on the detector. However, only the secondary charged particles of the gamma rays can produce Cherenkov light in the Cherenkov detector. The neutron and gamma-ray fluxes can be calculated by measuring the fluorescence and Cherenkov light. The GEANT4 Monte Carlo simulation toolkit is used to simulate the whole process occurring in the detector, whose optimum parameters are known. Analysis of the simulation results leads to a calculation method of neutron flux. This method is verified by calculating the neutron fluxes using pulsed n–γ mixed fields with different n/γ ratios, and the results show that the relative errors of all calculations are <5%. - Highlights: • A neutron detector is developed to discriminate 14-MeV fast neutrons and gamma rays. • The GEANT4 is used to optimize the parameters of the detector. • A calculation method of neutron flux is established through the simulation. • Several n/γ mixture fields are simulated to validate of the calculation method.

  14. Materials testing by computerized tomography with neutrons and gamma-rays

    Energy Technology Data Exchange (ETDEWEB)

    El-Ghobary, A M; Bakkoush, F A; Megahid, R M [Reactor and Neutron Physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)

    1997-12-31

    The method of computerized tomography by fast neutrons and gamma-rays are used for inspecting and testing of materials by non-destructive technique. The transmission technique was applied using narrow collimated beams of reactor neutrons and gamma-ray. The neutron and gamma-rays transmitted through the object inspection were measured by means of a neutron gamma detector with Ne - 213 liquid organic scintillator. The undesired pulses of neutrons or gamma-rays are rejected from the transmitted beam by a discrimination technique based on the difference in the decay part of light pulse produced by recoil electrons or recoil protons. The transmitted neutrons or gamma-rays for different projections used to get the image of the section through the object investigated using the method of filtered back projection (FBP) algorithm. 8 figs.

  15. Detection of fast neutrons in a plastic scintillator using digital pulse processing to reject gammas

    International Nuclear Information System (INIS)

    Reeder, P.L.; Peurrung, A.J.; Hansen, R.R.; Stromswold, D.C.; Hensley, W.K.; Hubbard, C.W.

    1999-01-01

    We report on neutron-gamma discrimination in a plastic scintillator based on the time delay inherent in second and third chance neutron scattering. Because of the time delay (∼3 ns) between the first and second scattering of a neutron, calculations of gammas and neutrons in a plastic scintillator predict that a neutron signal should be significantly broader than a pulse from a gamma event. Experimentally, we have used a fast digital oscilloscope coupled to a computer to examine individual pulses from neutron or gamma induced signals in fast scintillators coupled to a fast PMT. Individual neutron-induced signals were consistent with the predictions of our model, but gamma pulses were broader than expected. We present various tests to understand this phenomenon and discuss a way to overcome this problem

  16. Neutron Capture Gamma-Ray Spectroscopy. Proceedings of the International Symposium on Neutron Capture Gamma-Ray Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-11-15

    Experimental capabilities in the field of neutron capture gamma-ray spectroscopy have expanded greatly in the last few years; this has been due in large part to the advent of high-quality Ge(Li) detectors, improvements in electronic data processing, and improvements in bent-crystal spectrometers. Previously unsuspected phenomena, such as the '5. 5-MeV1 anomaly, have appeared and new research tools, such as neutron guide tubes, have been brought into use. Equally exciting developments have occurred in the theory of neutron capture. Complex spectra have yielded to analysis after account had been taken of such effects as vibration, rotation and Coriolis forces, and the theoretical prediction of capture spectra seems to be a future possibility. In view of the International Atomic Energy Agency's close interest in this subject and the need for an international exchange of ideas to analyse and study the latest developments, the organizers of the Symposium felt that work on neutron capture gamma-ray spectroscopy had achieved such valuable and significant results that the time had come for this information to be presented, examined and discussed internationally.

  17. Neutron Capture Gamma-Ray Spectroscopy. Proceedings of the International Symposium on Neutron Capture Gamma-Ray Spectroscopy

    International Nuclear Information System (INIS)

    1969-01-01

    Experimental capabilities in the field of neutron capture gamma-ray spectroscopy have expanded greatly in the last few years; this has been due in large part to the advent of high-quality Ge(Li) detectors, improvements in electronic data processing, and improvements in bent-crystal spectrometers. Previously unsuspected phenomena, such as the '5. 5-MeV1 anomaly, have appeared and new research tools, such as neutron guide tubes, have been brought into use. Equally exciting developments have occurred in the theory of neutron capture. Complex spectra have yielded to analysis after account had been taken of such effects as vibration, rotation and Coriolis forces, and the theoretical prediction of capture spectra seems to be a future possibility. In view of the International Atomic Energy Agency's close interest in this subject and the need for an international exchange of ideas to analyse and study the latest developments, the organizers of the Symposium felt that work on neutron capture gamma-ray spectroscopy had achieved such valuable and significant results that the time had come for this information to be presented, examined and discussed internationally

  18. Design of a versatile detector for the detection of charged particles, neutrons and gamma rays. Neutron interaction with the matter; Diseno de un detector versatil para la deteccion de particulas cargadas, neutrones y rayos gamma. Interaccion neutronica con la materia

    Energy Technology Data Exchange (ETDEWEB)

    Perez P, J J [Comision Nacional de Seguridad Nuclear y Salvaguardias, Mexico, D.F. (Mexico)

    1991-07-01

    The Fostron detector detects charged particles, neutrons and gamma rays with a reasonable discrimination power. Because the typical detectors for neutrons present a great uncertainty in the detection, this work was focused mainly to the neutron detection in presence of gamma radiation. Also there are mentioned the advantages and disadvantages of the Fostron detector.

  19. Some neutron and gamma radiation characteristics of plutonium cermet fuel for isotopic power sources

    Science.gov (United States)

    Neff, R. A.; Anderson, M. E.; Campbell, A. R.; Haas, F. X.

    1972-01-01

    Gamma and neutron measurements on various types of plutonium sources are presented in order to show the effects of O-17, O-18 F-19, Pu-236, age of the fuel, and size of the source on the gamma and neutron spectra. Analysis of the radiation measurements shows that fluorine is the main contributor to the neutron yields from present plutonium-molybdenum cermet fuel, while both fluorine and Pu-236 daughters contribute significantly to the gamma ray intensities.

  20. Discrimination of neutrons and gamma quanta with the aid of their power density spectra

    International Nuclear Information System (INIS)

    Buchmueller, R.

    1977-01-01

    The paper introduces a method of using only one fission chamber to discriminate the neutron flux against the gamma flux. The gamma chamber current may be several orders of magnitude higher than the neutron chamber current. In specially dimensioned fission chambers the neutrons and gamma quanta are made to generate different current pulses. Discrimination becomes possible by recording the power density spectrum of the mixture of pulses over a broad frequency range ( [de

  1. Evaluation of the neutron and gamma-ray production cross-sections for 55Mn

    International Nuclear Information System (INIS)

    Takahashi, H.

    1974-11-01

    The evaluation of neutron and gamma production cross sections for manganese-55 from 1.0 (10) -5 eV to 20.0 MeV for ENDF/ B-IV is summarized. Included are resonance parameters, neutron cross sections, angular and energy distribution of secondary neutrons, gamma multiplicities and transition probability array, gamma angular and energy distributions, nuclear model calculations, uncertainty estimates of cross sections, and evaluated cross sections. (U.S.)

  2. Development of criticality accident detector measuring neutrons and gamma-rays

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Ishii, Masato

    2005-01-01

    The authors developed a new criticality accident detector measuring neutrons and gamma-rays. The detector is a cylindrical plastic scintillator coupled to a current-mode operated photomultiplier, and is covered by an inner cadmium shell, acting as a neutron to gamma-ray converter, and a 5cm thick outer polyethylene moderator in order to respond to the same threshold triggering dose regardless of whether it was exposed to neutrons, gamma-rays or a mixture of the two radiations. (author)

  3. Neutron Assay System for Con?nement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Waste will be removed from confinement vessels remaining from 1970s-era experiments. Los Alamos has 9+ spherical confinement vessels remaining from experiments. Each vessel contains ∼ 500 lbs of radioactive debris such as actinide metals and oxides, metals, powdered silica, graphite, and wires and hardware. In order to dispose of the vessels, debris and contamination must be removed. Neutron assay system was designed to assay vessels before and after cleanout. System requirements are: (1) Modular and moveable; (2) Capable of detecting ∼100g 239 Pu equivalent in a 2-inch thick steel sphere with 6 foot diameter; and (3) Capable of safeguards-quality assays. Initial design parameters arethe use of 4-atm 3 He tubes with length of 6 feet, and 3 He tubes embedded in polyethelene for moderation. This paper describes the calibration of the Confinement Vessel Assay System (CVAS) and quantification of its uncertainties. Assay uncertainty depends on five factors: (1) Statistical uncertainty in the assay measurement; (2) Statistical uncertainty in the background measurement; (3) Statistical uncertainty in the isotopics determination - This should be much smaller than the other uncertainties; (4) Systematic uncertainty due to position bias; and (5) Systematic uncertainty due to fluctuations in cosmic ray spallation. This one can be virtually eliminated by performing the background measurement with an empty vessel - but that may not be possible. We used modeling and experiments to quantify the systematic uncertainties. The calibration assumes a uniform distribution of material, but reality will be different. MCNPX modeling was used to quantify the positional bias. The model was benchmarked to build confidence in its results. Material at top of vessel is 44% greater than amount assayed, according to singles. Material near 19-tube detector is 38% less than amount assayed, according to singles. Cosmic ray spallation contributes significantly to the background. Comparing rates

  4. Study of SMM flares in gamma-rays and neutrons

    Science.gov (United States)

    Dunphy, Philip P.; Chupp, Edward L.

    1992-01-01

    This report summarizes the results of the research supported by NASA grant NAGW-2755 and lists the papers and publications produced through the grant. The objective of the work was to study solar flares that produced observable signals from high-energy (greater than 10 MeV) gamma-rays and neutrons in the Solar Maximum Mission (SMM) Gamma-Ray Spectrometer (GRS). In 3 of 4 flares that had been studied previously, most of the neutrons and neutral pions appear to have been produced after the 'main' impulsive phase as determined from hard x-rays and gamma-rays. We, therefore, proposed to analyze the timing of the high-energy radiation, and its implications for the acceleration, trapping, and transport of flare particles. It was equally important to characterize the spectral shapes of the interacting energetic electrons and protons - another key factor in constraining possible particle acceleration mechanisms. In section 2.0, we discuss the goals of the research. In section 3.0, we summarize the results of the research. In section 4.0, we list the papers and publications produced under the grant. Preprints or reprints of the publications are attached as appendices.

  5. Neutron and gamma-ray spectra of 239PuBe and 241AmBe

    International Nuclear Information System (INIS)

    Vega-Carrillo, H.R.; Manzanares-Acuna, Eduardo; Becerra-Ferreiro, A.M.; Carrillo-Nunez, Aureliano

    2002-01-01

    Neutron and gamma-ray spectra of 239 PuBe and 241 AmBe were measured and their dosimetric features were calculated. Neutron spectra were measured using a multisphere neutron spectrometer with a 6 LiI(Eu) scintillator. The 239 PuBe neutron spectrum was measured in an open environment, while the 241 AmBe neutron spectrum was measured in a closed environment. Gamma-ray spectra were measured using a NaI(Tl) scintillator using the same experimental conditions for both sources. The effect of measuring conditions for the 241 AmBe neutron spectrum indicates the presence of epithermal and thermal neutrons. The low-resolution neutron spectra obtained with the multisphere spectrometer allows one to calculate the dosimetric features of neutron sources. At 100 cm both sources produce approximately the same count rate as that of the 4.4 MeV gamma-ray per unit of alpha emitter activity

  6. Gamma aminobutyric acid radioreceptor assay: a confirmatory quantitative assay for toxaphene in environmental and biological samples

    International Nuclear Information System (INIS)

    Saleh, M.A.; Blancato, J.N.

    1993-01-01

    Toxaphene is a complex mixture of polychlorinated monoterpenes, and was found to be acutely and chronically toxic to aquatic and wild life and posed a carcinogenic risk to humans before its ban in 1982. However, it is still found in the environment due to its relative persistence with an estimated half life time of about 10 years in soils. Toxaphenes neurotoxicity is attributed to a few isomers with a mode of action through binding to the chloride channel of the gamma-aminobutyric acid (GABA) receptor ionophore complex. [ 35 S] tertiary butylbicyclophosphorothionate (TBPS) with specific activity higher than 60 Ci/mmole has a high binding affinity to the same sites and is now commercially available and can be used to label the GABA receptor for the development of radioreceptor assay technique. The GABA receptor was prepared by a sequence of ultra centrifugation and dialysis of mammalian (rats, cows, catfish and goats) brain homogenates. The receptor is then labeled with [ 35 S] TBPS and the assay was conducted by measuring the displacement of radioactivity following incubation with the sample containing the analytes. The assay is fast, sensitive and requires very little or no sample preparation prior to the analysis. (Author)

  7. Slow neutrons and secondary gamma ray distributions in concrete shields followed by reflecting layers

    International Nuclear Information System (INIS)

    Makarious, A.S.; Swilem, Y.I.; Awwad, Z.; Bayomy, T.

    1993-01-01

    Slow neutrons and secondary gamma ray distributions in concrete shields with and without a reflecting layer behind layer behind the concrete shield have been investigated first in case of using a bare reactor beam and then on using a B-4 C filtered beam. The total and capture secondary gamma ray coefficient (B gamma and B gamma C ), the ratio of the reflected thermal neutron (gamma) the ratio of the secondary gamma rays caused by reflected neutrons to those caused transmitted neutrons (Th I gamma/F I gamma) and the effect of inserting a blocking layer (a B-4 C layer) between the concrete shield and the reflector on the suppression of the produced secondary gamma rays have been investigated. It was found that the presence of the reflector layer behind the concrete shield reflects some thermal neutrons back to the concrete shields and so it increases the number of thermal neutrons at the interface between the concrete shield and the reflector. Also the capture secondary gamma rays was increased at the interface between the two medii due to the capture of the reflected thermal neutrons in the concrete shields. It was shown that B-gamma is higher than and that B g amma B gamma C and I gamma T h/ I gamma i f for the different concrete types is higher in case of using the graphite reflector than that in using either water or paraffin reflectors. Putting a blocking layer (B 4 C layer) between the concrete shield and the reflector decreases the produced secondary gamma rays due to the absorption of the reflected thermal neutrons. 17 figs

  8. Gamma-neutron activation experiments using laser wakefield accelerators

    International Nuclear Information System (INIS)

    Leemans, W.P.; Rodgers, D.; Catravas, P.E.; Geddes, C.G.R.; Fubiani, G.; Esarey, E.; Shadwick, B.A.; Donahue, R.; Smith, A.

    2001-01-01

    Gamma-neutron activation experiments have been performed with relativistic electron beams produced by a laser wakefield accelerator. The electron beams were produced by tightly focusing (spot diameter ≅6 μm) a high power (up to 10 TW), ultra-short (≥50 fs) laser beam from a high repetition rate (10 Hz) Ti:sapphire (0.8 μm) laser system, onto a high density (>10 19 cm -3 ) pulsed gasjet of length ≅1.5 mm. Nuclear activation measurements in lead and copper targets indicate the production of electrons with energy in excess of 25 MeV. This result was confirmed by electron distribution measurements using a bending magnet spectrometer. Measured γ-ray and neutron yields are also found to be in reasonable agreement with simulations using a Monte Carlo transport code

  9. Gamma-burst emission from neutron-star accretion

    Science.gov (United States)

    Colgate, S. A.; Petschek, A. G.; Sarracino, R.

    1983-01-01

    A model for emission of the hard photons of gamma bursts is presented. The model assumes accretion at nearly the Eddington limited rate onto a neutron star without a magnetic field. Initially soft photons are heated as they are compressed between the accreting matter and the star. A large electric field due to relatively small charge separation is required to drag electrons into the star with the nuclei against the flux of photons leaking out through the accreting matter. The photon number is not increased substantially by Bremsstrahlung or any other process. It is suggested that instability in an accretion disc might provide the infalling matter required.

  10. Gamma and neutron irradiation tests on commercial IC op amps

    International Nuclear Information System (INIS)

    Kennedy, E.J.; Morris, A.C. Jr.; Su, D.K.

    1985-01-01

    Experimental results of gamma and neutron irradiation tests on 30 types of integrated-circuit operational amplifiers from 11 manufacturers are presented. All units were low-cost, commercial-grade devices. Op amps were evaluated for changes in offset voltage, input bias current, power supply current, open-loop gain, gain-bandwidth product, slew rate, power-supply and common-mode rejection ratios. Bipolar transistor op amps with resistive collector load resistors for the input stage indicated the best radiation hardness

  11. Neutron-Activated Gamma-Emission: Technology Review

    Science.gov (United States)

    2012-01-01

    in Be9 + α  C12 + n and Be9 + α  3He4 + n. Chadwick (5) made use of the naturally occurring α-emitter polonium - 210 , which decays to lead-206 with...emission, the variation of gamma attenuation with distance and the presence of organic clutter (in food , fertilizer, dirt road, etc.) makes it 8...a neutron source by mixing a radioisotope that emits alpha particles, such as radium or polonium , with a low atomic weight isotope, usually in the

  12. New detectors of neutron, gamma- and X-radiations

    CERN Document Server

    Lobanov, N S

    2002-01-01

    Paper presents new detectors to record absorbed doses of neutron, gamma- and X-ray radiations within 0-1500 Mrad range. DBF dosimeter is based on dibutyl phthalate. EDS dosimeter is based on epoxy (epoxide) resin, while SD 5-40 detector is based on a mixture of dibutyl phthalate and epoxy resin. Paper describes experimental techniques to calibrate and interprets the measurement results of absorbed doses for all detectors. All three detectors cover 0-30000 Mrad measured does range. The accuracy of measurements is +- 10% independent (practically) of irradiation dose rates within 20-2000 rad/s limits under 20-80 deg C temperature

  13. Time-of-flight spectrometer for the measurement of gamma correlated neutron spectra

    International Nuclear Information System (INIS)

    Andriashin, A.V.; Devkin, B.V.; Lychagin, A.A.; Minko, J.V.; Mironov, A.N.; Nesterenko, V.S.; Sztaricskai, T.; Petoe, G.; Vasvary, L.

    1986-01-01

    A time-of-flight spectrometer for the measurement of gamma correlated neutron spectra from (n,xnγ) reactions is described. The operation and the main parameters are discussed. The resolution in the neutron channel is 2.2 ns/m at the 150 keV neutron energy threshold. A simultaneous measurement of the time-of-flight and amplitude distributions makes it possible to study gamma correlated neutron spectra as well as the prompt gamma spectra in coincidence with selected energy neutrons. In order to test the spectrometer, measurements of the neutron spectrum in coincidence with the 846 keV gamma line of 56 Fe were carried out at an incident neutron energy of 14.1 MeV. (Auth.)

  14. Time-of-flight spectrometer for the measurement of gamma correlated neutron spectra

    International Nuclear Information System (INIS)

    Andryashin, A.V.; Devlein, B.V.; Lychagin, A.A.; Minko, Y.V.; Mironov, A.N.; Nesterenko, V.S.

    1986-01-01

    A time-of-flight spectrometer for the measurement of gamma correlated neutron spectra form (n,xnγ) reactions is described. The operation and the main parameters are discussed. The resolution in the neutron channel is 2.2 ns/m at the 150 keV neutron energy threshold. A simultaneous measurement of the time-of-flight and amplitude distributions makes it possible to study gamma correlated neutron spectra as well as the prompt gamma spectra in coincidence with selected energy neutrons. In order to test the spectrometer, measurements of the neutron spectrum in coincidence with the 846 keV gamma line of 56 Fe were carried out at an incident neutron energy of 14.1 MeV. (author). 3 figs., 6 refs

  15. Method and apparatus for neutron induced gamma ray logging for lithology identification

    International Nuclear Information System (INIS)

    Oliver, D.W.; Culver, R.B.

    1981-01-01

    The patent describes a neutron-gamma well logging technique which can distinguish between sandstone and limestone formations irrespective of water salinity in the formation. The formation surrounding a borehole is irradiated by fast neutrons and the resulting gamma rays are counted. The gamma rays are converted to electrical signals in three distinct steps; the first two signals result from gamma rays associated with calcium content of the formation and the third signal from gamma rays associated with silicon content. Gamma rays resulting from irradiation of calcium are counted at two non-contiguous energy bands. (O.T.)

  16. A novel detector assembly for detecting thermal neutrons, fast neutrons and gamma rays

    Energy Technology Data Exchange (ETDEWEB)

    Cester, D., E-mail: davide.cester@gmail.com [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Lunardon, M.; Moretto, S. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Nebbia, G. [INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Pino, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Sajo-Bohus, L. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Laboratorio de Fisica Nuclear, Universidad Simon Bolivar, Apartado 89000, 1080 A Caracas (Venezuela, Bolivarian Republic of); Stevanato, L.; Bonesso, I.; Turato, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy)

    2016-09-11

    A new composite detector has been developed by combining two different commercial scintillators. The device has the capability to detect gamma rays as well as thermal and fast neutrons; the signal discrimination between the three types is performed on-line by means of waveform digitizers and PSD algorithms. This work describes the assembled detector and its discrimination performance to be employed in the applied field.

  17. A novel detector assembly for detecting thermal neutrons, fast neutrons and gamma rays

    International Nuclear Information System (INIS)

    Cester, D.; Lunardon, M.; Moretto, S.; Nebbia, G.; Pino, F.; Sajo-Bohus, L.; Stevanato, L.; Bonesso, I.; Turato, F.

    2016-01-01

    A new composite detector has been developed by combining two different commercial scintillators. The device has the capability to detect gamma rays as well as thermal and fast neutrons; the signal discrimination between the three types is performed on-line by means of waveform digitizers and PSD algorithms. This work describes the assembled detector and its discrimination performance to be employed in the applied field.

  18. Gamma Interferon Release Assays for Detection of Mycobacterium tuberculosis Infection

    Science.gov (United States)

    Denkinger, Claudia M.; Kik, Sandra V.; Rangaka, Molebogeng X.; Zwerling, Alice; Oxlade, Olivia; Metcalfe, John Z.; Cattamanchi, Adithya; Dowdy, David W.; Dheda, Keertan; Banaei, Niaz

    2014-01-01

    SUMMARY Identification and treatment of latent tuberculosis infection (LTBI) can substantially reduce the risk of developing active disease. However, there is no diagnostic gold standard for LTBI. Two tests are available for identification of LTBI: the tuberculin skin test (TST) and the gamma interferon (IFN-γ) release assay (IGRA). Evidence suggests that both TST and IGRA are acceptable but imperfect tests. They represent indirect markers of Mycobacterium tuberculosis exposure and indicate a cellular immune response to M. tuberculosis. Neither test can accurately differentiate between LTBI and active TB, distinguish reactivation from reinfection, or resolve the various stages within the spectrum of M. tuberculosis infection. Both TST and IGRA have reduced sensitivity in immunocompromised patients and have low predictive value for progression to active TB. To maximize the positive predictive value of existing tests, LTBI screening should be reserved for those who are at sufficiently high risk of progressing to disease. Such high-risk individuals may be identifiable by using multivariable risk prediction models that incorporate test results with risk factors and using serial testing to resolve underlying phenotypes. In the longer term, basic research is necessary to identify highly predictive biomarkers. PMID:24396134

  19. Studies on bystander effects of 14MeV neutrons in human blood lymphocytes using CBMN assay

    International Nuclear Information System (INIS)

    Bakkiam, D.; Arul Anantha Kumar, A.; Sonwani, Swetha; Alaguraja, E.; Mathiyarasu, R.; Baskaran, R.; Venkatraman, B.

    2018-01-01

    Radiation induced Bystander Effects (RIBE) in cells generally describes the phenomenon that non-irradiated cells respond as if they have themselves been irradiated upon receiving signals from directly irradiated cells, either through partnering or medium transfer. While it has been well established that bystander effects could be induced by gamma radiation and alpha-particle radiation it is still a question whether neutrons induce bystander effects or not. In view of this, experiments were carried out to quantify cytogenetic damage in human blood lymphocytes induced by neutron directly and indirectly i.e. RIBE through medium transfer method. Cytokinesis Blocked MicroNucleus (CBMN) assay was used to study DNA damage events wherein micronuclei (MN) were scored in binucleated cells. Results of MN frequency in neutron direct and indirect irradiated blood lymphocytes (bystander samples) are compared

  20. Monitoring of plutonium contaminated solid waste streams. Chapter IV: Passive neutron assay

    International Nuclear Information System (INIS)

    Birkhoff, G.; Bondar, L.

    1978-01-01

    The fundamentals of the passive neutron technique for the non destructive assay of plutonium bearing materials are summarized. A reference monitor for the passive neutron assay of Pu contaminated solids is described in terms of instrumental design principles and performances. The theoretical model of this reference monitor with pertinent nuclear data and functions for the interpretation of experimental data is given

  1. Hematologic status of mice submitted to sublethal total body irradiation with mixed neutron-gamma radiation

    International Nuclear Information System (INIS)

    Herodin, F.; Court, L.

    1989-01-01

    The hematologic status of mice exposed to sublethal whole body irradiation with mixed neutron-gamma radiation (mainly neutrons) is studied. A slight decrease of the blood cell count is still observed below 1 Gy. The recovery of bone marrow granulocyte-macrophage progenitors seems to require more time than after pure gamma irradiation [fr

  2. Scanning of Cargo Containers by Gamma-Ray and Fast Neutron Radiography

    International Nuclear Information System (INIS)

    Yousri, A.M.; Bashter, I.I.; Megahid, M.R.; Osman, A.M.; Kansouh, W.A.; Reda, A.M.

    2011-01-01

    This paper describes the combined systems which were installed and tested to detect contraband smuggled in cargo containers. These combined systems are based on radiographers work by gamma-rays emitted from point source 60 Co with 0.5 Ci activity and neutrons emitted from point isotopic sources of Pu-α-Be as well as 14 MeV neutrons emitted from sealed tube neutron generator. The transmitted gamma ray through the inspected object was measured by gamma detection system with NaI(Tl) detector while the transmitted fast neutron beam was measured by a neutron gamma detection system with stilbene organic scintillator. The later possess the capability of discrimination between between gamma and neutron pulses using a discrimination system based on pulse shape discrimination method. The measured intensities of primary incident and transmitted beams of gamma-rays and fast neutrons were used to construct 2D cross-sectional images of the inspected objects hidden directly within benign materials of the container and for object screened by high dense material to stop object detection by gamma or X-rays. The constructed images for the inspected objects show the good capability and effectiveness of the installed gamma and neutron radiographers to detect illicit materials hidden in air cargo containers and sea containers of med size. They have also indicated that the developed scanning systems possess the ease of mobility and low cost of scanning

  3. Neutron interrogation system using high gamma ray signature to detect contraband special nuclear materials in cargo

    Science.gov (United States)

    Slaughter, Dennis R [Oakland, CA; Pohl, Bertram A [Berkeley, CA; Dougan, Arden D [San Ramon, CA; Bernstein, Adam [Palo Alto, CA; Prussin, Stanley G [Kensington, CA; Norman, Eric B [Oakland, CA

    2008-04-15

    A system for inspecting cargo for the presence of special nuclear material. The cargo is irradiated with neutrons. The neutrons produce fission products in the special nuclear material which generate gamma rays. The gamma rays are detecting indicating the presence of the special nuclear material.

  4. FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation

    International Nuclear Information System (INIS)

    Taormina, A.

    1978-05-01

    FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)

  5. Deficiency in Monte Carlo simulations of coupled neutron-gamma-ray fields

    NARCIS (Netherlands)

    Maleka, Peane P.; Maucec, Marko; de Meijer, Robert J.

    2011-01-01

    The deficiency in Monte Carlo simulations of coupled neutron-gamma-ray field was investigated by benchmarking two simulation codes with experimental data. Simulations showed better correspondence with the experimental data for gamma-ray transport only. In simulations, the neutron interactions with

  6. First results on irradiation of ceramic parallel plate chambers with gammas and neutrons

    International Nuclear Information System (INIS)

    Arefiev, A.; Bencze, Gy.L.; Bizzeti, A.; Choumilov, E.; Civinini, C.; Dajko, G.; D'Alessandro, R.; Fenyvesi, A.; Ferrando, A.; Fouz, M.C.; Iglesias, A.; Ivochkin, V.; Josa, M.I.; Malinin, A.; Meschini, M.; Molnar, J.; Pojidaev, V.; Salicio, J.M.; Tanko, L.; Vesztergombi, G.

    1996-01-01

    Ceramic parallel plate chambers were irradiated with gamma rays and neutrons. Results on radiation resistance are presented after 60 Mrad gamma and 0.5.10 16 neutrons per cm 2 irradiation of the detector surface. Results of activation analysis of chambers made of two different ceramic materials are also presented. (orig.)

  7. Burn-up measurements coupling gamma spectrometry and neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

    2006-07-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  8. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

    2006-01-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  9. DNA-repair after irradiation of cells with gamma-rays and neutrons

    International Nuclear Information System (INIS)

    Altmann, H.

    1975-11-01

    The structural alterations of calf thymus DNA produced by neutron or gamma irradiation were observed by absorption spectra, sedimentation rate and viscosity measurements. Mixed neutron-gamma irradiation produced fewer single and double strand breaks compared with pure gamma irradiation. RBE-values for mixed neutron-gamma radiation were less than 1, and DNA damage decreased with increasing neutron dose rate. Repair processes of DNA occuring after irradiation were measured in mouse spleen suspensions and human lymphocytes using autoradiographic methods and gradient centrifugations. The number of labelled cells was smaller after mixed neutron-gamma irradiation than after gamma irradiation. The rejoining of strand breaks in alkaline and neutral sucrose was more efficient after gamma irradiation than after mixed neutron-gamma irradiation. Finally, the effect of detergents Tween 80 and Nonident P40 on unscheduled DNA synthesis was studied by autoradiography after mixed neutron-gamma irradiation (Dn=5 krad). The results showed that the DNA synthesis was inhibited by detergent solutions of 0.002%

  10. Simulation of neutrons and gamma pulse signal and research on the pulse shape discrimination technology

    International Nuclear Information System (INIS)

    Zuo Guangxia; He Bin; Xu Peng; Qiu Xiaolin; Ma Wenyan; Li Sufen

    2012-01-01

    In neutrons detection, it is important to discriminate the neutron signals from the gamma-ray background. In this article, simulation of neutrons and gamma pulse signals is developed based on the LabVIEW platform. Two digital algorithms of the charge comparison method and the pulse duration time method are realized using 10000 simulation signals. Experimental results show that neutron and gamma pulse signals can be discriminated by the two methods, and the pulse duration time method is better than the charge comparison method. (authors)

  11. Technical Aspect for Operating Portable Prompt Gamma Neutron Activation Analysis (PGNAA) on Terengganu Inscribed Stone

    International Nuclear Information System (INIS)

    Rasif Mohd Zain; Hearie Hassan; Roslan Yahya

    2015-01-01

    Prompt Gamma Neutron Activation analysis (PGNAA) is a type of neutron activation analysis which can determined element with nearly no gamma ray decay after being irradiated by neutron sourced. Thus, element that cannot be determined by the conventional NAA for example H, B, N, Si and Cd, can be determine by PGNAA. This paper focuses on the technical working procedure for operating portable PGNAA in field work. The device is designed as a portable non-destructive investigation tool applying an isotopic neutron source (Cf-252) and a gamma-ray spectroscopy system for in-situ investigation. The studied have been carried out on Terengganu inscribed stone at Terengganu State Museum. (author)

  12. GEANT4 simulation study of a gamma-ray detector for neutron resonance densitometry

    International Nuclear Information System (INIS)

    Tsuchiya, Harufumi; Harada, Hideo; Koizumi, Mitsuo; Kitatani, Fumito; Takamine, Jun; Kureta, Masatoshi; Iimura, Hideki

    2013-01-01

    A design study of a gamma-ray detector for neutron resonance densitometry was made with GEANT4. The neutron resonance densitometry, combining neutron resonance transmission analysis and neutron resonance capture analysis, is a non-destructive technique to measure amounts of nuclear materials in melted fuels of the Fukushima Daiichi nuclear power plants. In order to effectively quantify impurities in the melted fuels via prompt gamma-ray measurements, a gamma-ray detector for the neutron resonance densitometry consists of cylindrical and well type LaBr 3 scintillators. The present simulation showed that the proposed gamma-ray detector suffices to clearly detect the gamma rays emitted by 10 B(n, αγ) reaction in a high environmental background due to 137 Cs radioactivity with its Compton edge suppressed at a considerably small level. (author)

  13. Dosimetry techniques of thermal neutrons and {gamma} radiation in reactor cores; Techniques de dosimetrie des neutrons thermiques et du rayonnement {gamma} dans les piles

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, J; Draganic, I; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Chemical studies under radiation done in the reactor cores require to be followed by dosimetry. When the irradiations are done in the reflector, one can limit to the measure of the {gamma} and the neutron radiation. For the dosimetry of the {gamma} radiation, a dosimeter of ferrous sulfate is convenient until doses of about 10{sup 6} rep. The use of aired oxalic acid solutions permits to reach 10{sup 7} rep. The dosimetry of thermal neutrons has been made with solutions of cobalt sulphate or paper filter impregnated with this salt. The total chemical effect of the {gamma} and of the slow neutrons radiation is obtained with solutions of ferrous sulfate added with lithium sulphate. (M.B.) [French] Les etudes de chimie sous radiation faites dans les piles exigent d'etre suivies par dosimetrie. Lorsque les irradiations sont effectues dans le reflecteur, on peut se limiter a doser le rayonnement {gamma} et les neutrons. Pour la dosimetrie du rayonnement {gamma}, un dosimetre a sulfate ferreux convient jusqu'a des doses d'environ 10{sup 6} rep. L'emploi de solutions aerees d'acide oxalique permet d'atteindre 10{sup 7} rep. La dosimetrie des neutrons thermiques a ete faite avec des solutions de sulfate de cotalt ou du papier filtre impregne de ce sel. L'effet chimique total du rayonnement {gamma} et des neutrons lents est obtenu avec des solutions de sulfate ferreux additionne de sulfate de lithium. (M.B.)

  14. An investigation of the neutron die-away time in passive neutron waste assay systems

    International Nuclear Information System (INIS)

    Baeten, P.; Bruggeman, M.; Carchon, R.

    1997-02-01

    Neutron coincidence counting applied to the assay of Pu-bearing waste is commonly based on the assumption that the time intervals between detected fission neutrons are distributed according to a mono-exponential function, often called Rossi-alpha distribution. The time constant of this characteristic exponential function is generally referred to as the die-away time of the detector assembly. In fact, the distribution of time intervals is derived from the more fundamental arrival time distribution, which is also assumed to obey a mono-exponential law. In view of the design studies for a neutron counter, the validity of this basic assumption was investigated. Different parameters such as neutron moderation and absorption in the sample and the presence of cadmium-lining were investigated by means of Monte Carlo simulations using the NCNP-code. The simulation results lead to the conclusion that the description of the arrival time function with a mono-exponential function with a sample-independent die-away time is only a first approximations. The mono-exponential decay is perturbed by a second time component related to the detection of neutrons already thermalized in the sample. This thermal component cannot be described by a mono-exponential function, but has a characteristic shape with a fast build-up reaching a maximum followed by a slow decay as a function of the arrival time. The relative contribution of this component strongly depends on the absorption and moderation of the sample matrix. This component cannot be described by a simple analytical expression involving sample related parameters. Hence, no direct useful information can be withdrawn from the arrival time probability function to characterize the waste matrix. The thermal component can be strongly suppressed by the use of cadmium-lining in front of the detector blocks simplifying the mathematical description of the arrival time probability function. Indications of the bias introduced by an inaccurate

  15. Neutron and gamma dose and spectra measurements on the Little Boy replica

    International Nuclear Information System (INIS)

    Hoots, S.; Wadsworth, D.

    1984-01-01

    The radiation-measurement team of the Weapons Engineering Division at Lawrence Livermore National Laboratory (LLNL) measured neutron and gamma dose and spectra on the Little Boy replica at Los Alamos National Laboratory (LANL) in April 1983. This assembly is a replica of the gun-type atomic bomb exploded over Hiroshima in 1945. These measurements support the National Academy of Sciences Program to reassess the radiation doses due to atomic bomb explosions in Japan. Specifically, the following types of information were important: neutron spectra as a function of geometry, gamma to neutron dose ratios out to 1.5 km, and neutron attenuation in the atmosphere. We measured neutron and gamma dose/fission from close-in to a kilometer out, and neutron and gamma spectra at 90 and 30 0 close-in. This paper describes these measurements and the results. 12 references, 13 figures, 5 tables

  16. ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

  17. The transport of neutrons and gamma-rays in the air

    International Nuclear Information System (INIS)

    Adamski, J.

    1980-01-01

    The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)

  18. Are 0.1%-accurate gamma-ray assays possible for 235U solutions

    International Nuclear Information System (INIS)

    Parker, J.L.

    1983-01-01

    The factors influencing the accuracy of passive gamma-ray assay of uniform, homogeneous solution samples have been studied in some detail, particularly for the assay of 235 U in uranium solutions. Factors considered are the overall long-term electronic stability, the information losses caused by the rate-related electronic processes of pulse pileup and dead-time, and the self-attenuation of gamma rays within the samples. Both experimental and computational studies indicate that gamma-ray assay procedures for solution samples of moderate size (from approx. 10 to perhaps a few hundred milliliters) are now capable of accuracies approaching 0.1% in many practical cases

  19. Observation of neutron standing waves at total reflection by precision gamma spectroscopy

    International Nuclear Information System (INIS)

    Aksenov, V.L.; Gundorin, N.A.; Nikitenko, Yu.V.; Popov, Yu.P.; Cser, L.

    1998-01-01

    Total reflection of polarized neutrons from the layered structure glass/Fe (1000 A Angstrom)/Gd (50 A Angstrom) is investigated by registering neutrons and gamma-quanta from thermal neutron capture. The polarization ratio of gamma counts of neutron beams polarized in and opposite the direction of the magnetic field is measured. The polarization ratio is larger than unity for the neutron wavelengths λ 2.2 A Angstrom. Such behaviour of the wavelength dependence of the gamma-quanta polarization ratio points to the fact that over the surface of the Fe Layer a neutron standing wave caused by the interference of the incident neutron wave and the wave refracted from the magnetized Fe layer is formed

  20. Time-of-flight discrimination between gamma-rays and neutrons by using artificial neural networks

    International Nuclear Information System (INIS)

    Akkoyun, S.

    2013-01-01

    Highlights: ► Time-of-flight (tof) is an obvious method for separation between gamma and neutron particles. ► tof distributions are obtained by neural networks. ► Neural network method is consistent with the experimental results. ► Neural networks can classify different events for discrimination. - Abstract: In gamma-ray spectroscopy, a number of neutrons are emitted from the nuclei together with the gamma-rays. These neutrons influence gamma-ray spectra. An obvious method for discrimination between neutrons and gamma-rays is based on the time-of-flight (tof) technique. In this work, the tof distributions of gamma-rays and neutrons were obtained both experimentally and by using artificial neural networks (ANNs). It was shown that, ANN can correctly classify gamma-ray and neutron events. Also, for highly nonlinear detector response for tof, we have constructed consistent empirical physical formulas (EPFs) by appropriate ANNs. These ANN–EPFs can be used to derive further physical functions which could be relevant to discrimination between gamma-rays and neutrons

  1. Gamma-neutron pagers make revolution in combating nuclear terrorism

    International Nuclear Information System (INIS)

    Stavrov, A.; Kagan, L.; Antonovski, A.

    2002-01-01

    Full text: After the events of September 11, 2001 it became quite evident that to combat terrorism, nuclear terrorism including, comprehensive measures able to prevent acts of terrorism are required. One of the important factors that ensure solving this task is technical means that allow early detection of tolls of terrorism. All the aforesaid is especially true for means of detection of radioactive materials. One of the most effective and convenient instruments designed for detection of radioactive materials, including nuclear materials, are gamma-neutron pagers that are instruments of the new generation developed in the last year. These pagers are sensitive, small and rugged monitors designed for detection of special nuclear materials and other radioactive materials by their gamma and/or neutron radiation. This detection is based on a comparison of the radiation intensity of an object and a background. These instruments will immediately inform an individual about the presence of a radioactive source by audible, visual, or vibration alarm. They are easy-to-operate instruments and a user does not need to be an expert in nuclear physics or radiation protection. A pager may be worn in pocket or on a belt ensuring the radiation protection of an individual. This instrument operates in automatic mode during a labor shift. It is provided with a non-volatile memory to store information about a history of operation: date and time when an instrument is turned on/off; the time and radiation levels, gamma and/or neutron, that exceed the alarm threshold; the current values of the count rate, etc. This information can be transmitted to a PC. These instruments can be used as the first level alarming devices at sites where installation of fixed monitors is impossible or inexpedient. Pagers are relatively inexpensive, sensitive and small instruments. Therefore it is advised that each person involved in control of goods transportation, vehicles and people uses such instruments. A

  2. The optimization of gamma spectra processing in prompt gamma neutron activation analysis (PGNAA)

    Energy Technology Data Exchange (ETDEWEB)

    Pinault, Jean-Louis [IAEA Expert, 96 rue du Port David, 45370 Dry (France)], E-mail: jeanlouis_pinault@hotmail.fr; Solis, Jose [Instituto Peruano de Energia Nuclear, Av. Canada No. 1470, San Borja, Lima 41 (Peru)

    2009-04-15

    The uncertainty of the elemental analysis is one of the major factors governing the utility of on-line Prompt Gamma Neutron Activation Analysis (PGNAA) in the blending and sorting of bulk materials. In this paper, a general method applicable to Gamma spectra processing is presented and applied to PGNAA in mineral industry. Based on the Fourier transform of spectra and their de-correlation in the Fourier space (the improvement of the conditioning of the correlation matrix), processing of overlapping of characteristic peaks minimizes the propagation of random errors, which optimizes the accuracy and decreases the detection limits of elemental analyses. In comparison with classical methods based on the linear combinations of relevant regions of spectra the improvement may be considerable, especially when several elements are interfering. The method is applied to four case stories covering both borehole logging and on-line analysis on conveyor belt of raw materials.

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

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

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

  4. Apparatus and method for identification of matrix materials in which transuranic elements are embedded using thermal neutron capture gamma-ray emission

    Science.gov (United States)

    Close, D.A.; Franks, L.A.; Kocimski, S.M.

    1984-08-16

    An invention is described that enables the quantitative simultaneous identification of the matrix materials in which fertile and fissile nuclides are embedded to be made along with the quantitative assay of the fertile and fissile materials. The invention also enables corrections for any absorption of neutrons by the matrix materials and by the measurement apparatus by the measurement of the prompt and delayed neutron flux emerging from a sample after the sample is interrogated by simultaneously applied neutrons and gamma radiation. High energy electrons are directed at a first target to produce gamma radiation. A second target receives the resulting pulsed gamma radiation and produces neutrons from the interaction with the gamma radiation. These neutrons are slowed by a moderator surrounding the sample and bathe the sample uniformly, generating second gamma radiation in the interaction. The gamma radiation is then resolved and quantitatively detected, providing a spectroscopic signature of the constituent elements contained in the matrix and in the materials within the vicinity of the sample. (LEW)

  5. Gamma/neutron competition above the neutron separation energy in delayed neutron emitters

    Directory of Open Access Journals (Sweden)

    Valencia E.

    2014-03-01

    Full Text Available To study the β-decay properties of some well known delayed neutron emitters an experiment was performed in 2009 at the IGISOL facility (University of Jyväskylä in Finland using Total Absorption γ-ray Spectroscopy (TAGS technique. The aim of these measurements is to obtain the full β-strength distribution below the neutron separation energy (Sn and the γ/neutron competition above. This information is a key parameter in nuclear technology applications as well as in nuclear astrophysics and nuclear structure. Preliminary results of the analysis show a significant γ-branching ratio above Sn.

  6. Detection of explosive substances by tomographic inspection using neutron and gamma-ray spectroscopy

    International Nuclear Information System (INIS)

    Farahmand, M.; Boston, A.J.; Grint, A.N.; Nolan, P.J.; Joyce, M.J.; Mackin, R.O.; D'Mellow, B.; Aspinall, M.; Peyton, A.J.; Silfhout, R. van

    2007-01-01

    In recent years the detection and identification of hazardous materials has become increasingly important. This work discusses research and development of a technique which is capable of detecting and imaging hidden explosives. It is proposed to utilise neutron interrogation of the substances under investigation facilitating the detection of emitted gamma radiation and scattered neutrons. Pulsed fast neutron techniques are attractive because they can be used to determine the concentrations of the light elements (hydrogen, carbon, nitrogen, and oxygen) which can be the primary components of explosive materials. Using segmented High Purity Ge (HPGe) detectors and digital pulse processing [R.J. Cooper, G. Turk, A.J. Boston, H.C. Boston, J.R. Cresswell, A.R. Mather, P.J. Nolan, C.J. Hall, I. Lazarus, J. Simpson, A. Berry, T. Beveridge, J. Gillam, R.A. Lewis, in: Proceedings of the 7th International Conference on Position Sensitive Detectors, Nuclear Instruments and Methods A, in press; I. Lazarus, D.E. Appelbe, A. J. Boston, P.J. Coleman-Smith, J.R. Cresswell, M. Descovich, S.A.A. Gros, M. Lauer, J. Norman, C.J. Pearson, V.F.E. Pucknell, J.A. Sampson, G. Turk, J.J. Valiente-Dobon, IEEE Trans. Nucl. Sci., 51 (2004) 1353; R.J. Cooper, A.J. Boston, H.C. Boston, J.R. Cresswell, A.N. Grint, A.R. Mather, P.J. Nolan, D.P. Scraggs, G. Turk, C.J. Hall, I. Lazarus, A. Berry, T. Beveridge, J. Gillam, R.A. Lewis, in: Proceedings of the 11th International Symposium on Radiation Measurements and Application, 2006. ] the scatter path of incident photons can be reconstructed to determine the origin of the gamma-rays without the need for mechanical collimation by applying the Compton camera principle [V. Schonfelder, A. Hirner, K. Schneider, Nucl. Instr. and Meth. 107 (1973) 385; R.W. Todd, J.M. Nightingale, D.B. Everett, Nature 251 (1974) 132. ]. In addition, it is proposed to utilise the scattered neutrons which recoil from the materials being assayed, detecting them with a fast

  7. Neutron and/or gamma radiation detecting system

    International Nuclear Information System (INIS)

    Cerff, K.

    1985-01-01

    A large reception surface for the radiation to be detected is formed on a body of scintillation material (ZnS-AG with B matrix) which is adapted to convert neutron or gamma radiation into light energy. A large number of fiber light conductors is embedded in the body of scintillation material such that the fibers extend essentially parallel and fully across the reception surface of the body of scintillation material. The light energy, upon propagation along the fiber light conductors, is coupled into the conductors along the surface of the fibers which are unisotropic. This arrangement permits the use of unisotropic light conductor systems which provide for a separation of light collecting and light transmitting functions which results in a substantial reduction of light absorption losses during light transmission so that most of the light energy coupled into the fiber light conductors reaches the optoelectronic amplifier coupled to the end of the light conductors. (orig./HP) [de

  8. Neutron and gamma ray streaming experiments at the fast neutron source reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Yanagisawa, Ichiro; Akiyama, Masatsugu; An, Shigehiro

    1979-07-01

    Neutron and gamma ray streaming experiments were performed in the ducts and cavities that were located in the heavy concrete shields of the fast neutron source reactor YAYOI of University of Tokyo. The configurations have the feature that the streaming through the ducts are occurred following the scattering in the cavity. The axes of the ducts are perpendicular to the source radiation from the core. The spectrum of the source was modified by putting a plug in the beam hole of the core. An aluminum plug and the plug which contains paraffin were used. The decay in the ducts, however, hardly depends on the source spectrum. The decay in the ducts is nearly exponential. (author)

  9. Optimized Design of Spacing in Pulsed Neutron Gamma Density Logging While Drilling

    Directory of Open Access Journals (Sweden)

    ZHANG Feng;HAN Zhong-yue;WU He;HAN Fei

    2016-10-01

    Full Text Available Radioactive source, used in traditional density logging, has great impact on the environment, while the pulsed neutron source applied in the logging tool is more safety and greener. In our country, the pulsed neutron-gamma density logging technology is still in the stage of development. Optimizing the parameters of neutron-gamma density instrument is essential to improve the measuring accuracy. This paper mainly studied the effects of spacing to typical neutron-gamma density logging tool which included one D-T neutron generator and two gamma scintillation detectors. The optimization of spacing were based on measuring sensitivity and counting statistic. The short spacing from 25 to 35 cm and long spacing from 60 to 65 cm were selected as the optimal position for near and far detector respectively. The result can provide theoretical support for design and manufacture of the instrument.

  10. Effect of gamma and neutron irradiation on the mechanical properties of Spectralon™ porous PTFE

    Energy Technology Data Exchange (ETDEWEB)

    Gourdin, William H., E-mail: gourdin1@llnl.gov [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA USA (United States); Datte, Philip; Jensen, Wayne; Khater, Hesham; Pearson, Mark [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA USA (United States); Girard, Sylvain [Laboratoire Hubert Curien − UMR CNRS 5516, 18 rue du Pr. Benoît Lauras, F-42000 Saint Etienne (France); Paillet, Philippe; Alozy, Eric [CEA, DAM, DIF, F-91297 Arpajon (France)

    2016-11-15

    Highlights: • The effects of neutrons and gammas on PTFE are equivalent for a given absorbed dose. • A neutron fluence of 10{sup 13} n/cm{sup 2} corresponds to a gamma dose of 200 Gy. • The dose-to-fluence conversion factor is approximately 5 × 10{sup 10} n/(cm{sup 2}-Gy). • Irradiation in a low-oxygen environment enhances loads and elongations. • Mechanical properties of PTFE will deteriorate at a neutron fluence of 10{sup 13} n/cm{sup 2}. - Abstract: We establish a correspondence between the mechanical properties (maximum load and failure elongation) of Spectralon™ porous PTFE irradiated with 14 MeV neutrons and 1.17 and 1.33 MeV gammas from a cobalt-60 source. From this correspondence we infer that the effects of neutrons and gammas on this material are approximately equivalent for a given absorbed dose.

  11. Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

    International Nuclear Information System (INIS)

    Genreith, Christoph

    2015-01-01

    Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237 Np, 241 Am and 242 Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237 Np were identified, as well as 19 of 241 Am, and 127 prompt γ-rays of 242 Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237 Np was observed at an energy of E γ =182.82(10) keV associated with a partial capture cross section of σ γ =22.06(39) b. The most intense prompt γ-ray lines of 241 Am and of 242 Pu were observed at E γ =154.72(7) keV with σ γ =72.80(252) b and E γ =287.69(8) keV with σ γ =7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237 Np, 241 Am and 242 Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was

  12. Thin film CdTe based neutron detectors with high thermal neutron efficiency and gamma rejection for security applications

    Energy Technology Data Exchange (ETDEWEB)

    Smith, L.; Murphy, J.W. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Kim, J. [Korean Research Institute of Standards and Science, Daejeon 305-600 (Korea, Republic of); Rozhdestvenskyy, S.; Mejia, I. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Park, H. [Korean Research Institute of Standards and Science, Daejeon 305-600 (Korea, Republic of); Allee, D.R. [Flexible Display Center, Arizona State University, Phoenix, AZ 85284 (United States); Quevedo-Lopez, M. [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States); Gnade, B., E-mail: beg031000@utdallas.edu [Materials Science and Engineering, University of Texas at Dallas, Richardson, TX 75080 (United States)

    2016-12-01

    Solid-state neutron detectors offer an alternative to {sup 3}He based detectors, but suffer from limited neutron efficiencies that make their use in security applications impractical. Solid-state neutron detectors based on single crystal silicon also have relatively high gamma-ray efficiencies that lead to false positives. Thin film polycrystalline CdTe based detectors require less complex processing with significantly lower gamma-ray efficiencies. Advanced geometries can also be implemented to achieve high thermal neutron efficiencies competitive with silicon based technology. This study evaluates these strategies by simulation and experimentation and demonstrates an approach to achieve >10% intrinsic efficiency with <10{sup −6} gamma-ray efficiency.

  13. Effect of Gamma Rays on Fast Neutron Registration in CR-39

    CERN Document Server

    Kobzev, A P; El-Halem, A A; Abdul-Ghaphar, U S; Salama, T A

    2002-01-01

    A set of CR-39 plastic detectors with front PE radiator was exposed to Am-Be neutron source, which has an emission rate of 0.86\\cdot 10^{7} sec^{-1}, and the neutron dose equivalent rate 1 m apart from the source is equal to 11 mrem/hr. Another set of samples was irradiated by a neutron dose of 4 rem, then exposed to different gamma-ray doses using ^{60}Co source. It was found that the track density grows with the increase of neutron dose and etching time. It was also found that the bulk etching rate V_{B}, the track diameter and the sensitivity of the CR-39 plastic detector with respect to the neutron irradiation increased with increasing gamma-ray dose in the range 1?10 Mrad. These results show that CR-39 can be considered as a promising fast neutron dosimeter and gamma-ray dosimeter.

  14. A compact neutron beam generator system designed for prompt gamma nuclear activation analysis.

    Science.gov (United States)

    Ghassoun, J; Mostacci, D

    2011-08-01

    In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5). 2010 Elsevier Ltd. All rights reserved.

  15. Attenuation of Reactor Gamma Radiation and Fast Neutrons Through Large Single-Crystal Materials

    International Nuclear Information System (INIS)

    Adib, M.

    2009-01-01

    A generalized formula is given which, for neutron energies in the range 10-4< E< 10 eV and gamma rays with average energy 2 MeV , permits calculation of the transmission properties of several single crystal materials important for neutron scattering instrumentation. A computer program Filter was developed which permits the calculation of attenuation of gamma radiation, nuclear capture, thermal diffuse and Bragg-scattering cross-sections as a function of materials constants, temperature and neutron energy. The applicability of the deduced formula along with the code checked from the obtained agreement between the calculated and experimental neutron transmission through various single-crystals A feasibility study for use of Si, Ge, Pb, Bi and sapphire is detailed in terms of optimum crystal thickness, mosaic spread and cutting plane for efficient transmission of thermal reactor neutrons and for rejection of the accompanying fast neutrons and gamma rays.

  16. Measurements of keV-neutron capture {gamma} rays of fission products. 3

    Energy Technology Data Exchange (ETDEWEB)

    Igashira, Masayuki [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1997-03-01

    {gamma} rays from the keV-neutron capture reactions by {sup 143,145}Nd and {sup 153}Eu have been measured in a neutron energy region of 10 to 80 keV, using a large anti-Compton NaI(Tl) {gamma}-ray spectrometer and the {sup 7}Li(p,n){sup 7}Be pulsed neutron source with a 3-MV Pelletron accelerator. The preliminary results for the capture cross sections and {gamma}-ray spectra of those nuclei are presented and discussed. (author)

  17. Using a Borated Panel to Form a Dual Neutron-Gamma Detector

    Energy Technology Data Exchange (ETDEWEB)

    Scott Wilde; Raymond Keegan

    2008-06-20

    A borated polyethylene plane placed between a neutron source and a gamma spectrometer is used to form a dual neutron-gamma detection system. The polyethylene thermalizes the source neutrons so that they are captured by {sup 10}B to produce a flux of 478 keV gamma-rays that radiate from the plane. This results in a buildup of count rate in the detector over that from a disk of the same diameter as the detector crystal (same thickness as the panel). Radiation portal systems are a potential application of this technique.

  18. Investigation on neutron/gamma discrimination phenomena in plastic scintillators

    International Nuclear Information System (INIS)

    Blanc, Pauline

    2014-01-01

    This PhD topic was born from misunderstandings and incomplete knowledge of the mechanism and relative effectiveness of neutron and gamma-ray (n/γ) discrimination between plastic scintillators compared to liquid scintillators. The shape of the light pulse these materials generate following interaction with an ionizing particle (predominantly recoil protons in the case of neutrons and electrons in the case of gamma-rays) is different in time in a way that depends on the detected particle (nature and energy). It is this fact that enables separation (PSD). The behavior in liquid scintillators has been extensively studied experimentally for practical applications. Only recently has it been shown that a weak separation can also be achieved using specially prepared plastics. The study of this system presents an open field and the understanding of both liquids and plastics with respect to their PSD properties is far from complete. This work is dedicated to exploring the fundamental photophysical phenomena at play in the generation of luminescence emission, following the interaction of ionizing radiation with organic scintillators. For this purpose, firstly a detailed literature review of the state-of-the-art has been conducted extending from 1960 to the present day. Secondly a complete characterization of the main scintillating materials has been conducted to define their fluorescence properties and the characteristics of their scintillation under irradiation. Thirdly a proton beam has been used to simulate recoil protons to quantify under controlled laboratory conditions their specific energy deposition in a plastic scintillator with PSD properties. The fourth part of this thesis is devoted to the study of PSD efficiency of scintillators as a function of their molecular structure. This investigation has led to a plastic scintillator prepared in our laboratory with good PSD properties and a patent submission. Finally, photophysical experiments were performed using a

  19. Gamma-ray bursts from fast, galactic neutron stars

    International Nuclear Information System (INIS)

    Colgate, S.A.; Leonard, P.J.

    1996-01-01

    What makes a Galactic model of gamma-ray bursts (GBs) feasible is the observation of a new population of objects, fast neutron stars, that are isotropic with respect to the galaxy following a finite period, ∼30 My, after their formation (1). Our Galactic model for the isotropic component of GBs is based upon high-velocity neutron stars (NSs) that have accretion disks. These fast NSs are formed in tidally locked binaries, producing a unique population of high velocity (approx-gt 10 3 kms -1 ) and slowly rotating (8 s) NSs. Tidal locking occurs due to the meridional circulation caused by the conservation of angular momentum of the tidal lobes. Following the collapse to a NS and the explosion, these lobes initially perturb the NS in the direction of the companion. Subsequent accretion (1 to 2 s) occurs on the rear side of the initial motion, resulting in a runaway acceleration of the NS by neutrino emission from the hot accreted matter. The recoil momentum of the relativistic neutrino emission from the localized, down flowing matter far exceeds the momentum drag of the accreted matter. The recoil of the NS is oriented towards the companion, but the NS misses because of the pre-explosion orbital motion. The near miss captures matter from the companion and forms a disk around the NS. Accretion onto the NS from this initially gaseous disk due to the ''alpha'' viscosity results in a soft gamma-ray repeater phase, which lasts ∼10 4 yr. Later, after the neutron star has moved ∼30 kpc from its birthplace, solid bodies form in the disk, and accrete to planetoid size bodies after ∼3x10 7 years. Some of these planetoid bodies, with a mass of ∼10 21 endash 10 22 g, are perturbed into an orbit inside the tidal distortion radius of approx-gt 10 5 km. Of these ∼1% are captured by the magnetic field of the NS at R 3 km to create GBs

  20. The comparison of four neutron sources for Prompt Gamma Neutron Activation Analysis (PGNAA) in vivo detections of boron.

    Science.gov (United States)

    Fantidis, J G; Nicolaou, G E; Potolias, C; Vordos, N; Bandekas, D V

    A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were carried out to compare the performance of the proposed PGNAA system using four different neutron sources ( 241 Am/Be, 252 Cf, 241 Am/B, and DT neutron generator). Among the different systems the 252 Cf neutron based PGNAA system has the best performance.

  1. Measurements of prompt gamma-rays from fast-neutron induced fission with the LICORNE directional neutron source

    CERN Document Server

    Wilson, J N; Halipre, P; Oberstedt, S; Oberstedt, A

    2014-01-01

    At the IPN Orsay we have developed a unique, directional, fast neutron source called LICORNE, intended initially to facilitate prompt fission gamma measurements. The ability of the IPN Orsay tandem accelerator to produce intense beams of $^7$Li is exploited to produce quasi-monoenergetic neutrons between 0.5 - 4 MeV using the p($^7$Li,$^7$Be)n inverse reaction. The available fluxes of up to 7 × 10$^7$ neutrons/second/steradian for the thickest hydrogen-rich targets are comparable to similar installations, but with two added advantages: (i) The kinematic focusing produces a natural neutron beam collimation which allows placement of gamma detectors adjacent to the irradiated sample unimpeded by source neutrons. (ii) The background of scattered neutrons in the experimental hall is drastically reduced. The dedicated neutron converter was commissioned in June 2013. Some preliminary results from the first experiment using the LICORNE neutron source at the IPN Orsay are presented. Prompt fission gamma rays from fas...

  2. Preliminary study about frequencies of unstable chromosome alterations induced by gamma beam and neutron-gamma mixed field

    International Nuclear Information System (INIS)

    Mendes, Mariana E.; Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

    2011-01-01

    The estimate on approximate dose in exposed individual can be made through conventional cytogenetic analysis of dicentric, this technique has been used to support physical dosimetry. It is important to estimate the absorbed dose in case of accidents with the aim of developing an appropriate treatment and biological dosimetry can be very useful in case where the dosimetry is unavailable. Exposure to gamma and neutron radiation leads to the same biological effects such as chromosomal alterations and cancer. However, neutrons cause more genetic damage, such as mutation or more structural damage, such as chromosome alterations. The aim of research is to compare frequencies of unstable chromosome alterations induced by a gamma beam with those from neutron-gamma mixed field. Two blood samples were obtained from one healthy donor and irradiated at different sources. The first sample was exposed to mixed field neutron-gamma sources 241 AmBe at the Neutron Calibration Laboratory (NCL - CRCN/NE - PE - Brazil) and the second one was exposed to 137 Cs gamma rays at 137 Cs Laboratory (CRCN/NE - PE - Brazil), both exposures resulting in an absorbed dose of 0.66Gy. Mitotic metaphase cells were obtained by lymphocyte culture for chromosomal analysis and slides were stained with Giemsa 5%. These preliminary results showed a similarity in associated dicentrics frequency per cell (0.041 and 0.048) after 137 Cs and 241 AmBe sources irradiations, respectively. However, it was not observed centric rings frequency per cell (0.0 and 0.027). This study will be continue to verify the frequencies of unstable chromosome alterations induced by only gamma beam and neutron-gamma mixed field. (author)

  3. Preliminary study about frequencies of unstable chromosome alterations induced by gamma beam and neutron-gamma mixed field

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, Mariana E.; Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Calixto, Merilane S.; Santos, Neide [Universidade Federal de Pernanmbuco (CCB/UFPE), Recife, PE (Brazil). Centro de Ciencias Biologicas. Dept. de Genetica

    2011-07-01

    The estimate on approximate dose in exposed individual can be made through conventional cytogenetic analysis of dicentric, this technique has been used to support physical dosimetry. It is important to estimate the absorbed dose in case of accidents with the aim of developing an appropriate treatment and biological dosimetry can be very useful in case where the dosimetry is unavailable. Exposure to gamma and neutron radiation leads to the same biological effects such as chromosomal alterations and cancer. However, neutrons cause more genetic damage, such as mutation or more structural damage, such as chromosome alterations. The aim of research is to compare frequencies of unstable chromosome alterations induced by a gamma beam with those from neutron-gamma mixed field. Two blood samples were obtained from one healthy donor and irradiated at different sources. The first sample was exposed to mixed field neutron-gamma sources {sup 241}AmBe at the Neutron Calibration Laboratory (NCL - CRCN/NE - PE - Brazil) and the second one was exposed to {sup 137}Cs gamma rays at {sup 137}Cs Laboratory (CRCN/NE - PE - Brazil), both exposures resulting in an absorbed dose of 0.66Gy. Mitotic metaphase cells were obtained by lymphocyte culture for chromosomal analysis and slides were stained with Giemsa 5%. These preliminary results showed a similarity in associated dicentrics frequency per cell (0.041 and 0.048) after {sup 137}Cs and {sup 241}AmBe sources irradiations, respectively. However, it was not observed centric rings frequency per cell (0.0 and 0.027). This study will be continue to verify the frequencies of unstable chromosome alterations induced by only gamma beam and neutron-gamma mixed field. (author)

  4. The progress of neutron induced prompt gamma analysis technique in 1988-2002

    International Nuclear Information System (INIS)

    Liu Yuren; Jing Shiwei

    2003-01-01

    The new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988-2002 are described. The pulse fast-thermal neutron activation analysis method, which utilizes the inelastic reaction and capture reaction jointly is employed to measure the elemental content in the material more efficiently. The lifetime of the neutron generator is more than 10000 h and the capability of HPGe, TeZeCd and MCA (multi-channel analyser) reaches the high level. At the same time, Monte Carlo Library least-square method is used to solve the nonlinearity problem in the PGNAA (Prompt Gamma Neutron Activation Analysis)

  5. ICF ignition capsule neutron, gamma ray, and high energy x-ray images

    Science.gov (United States)

    Bradley, P. A.; Wilson, D. C.; Swenson, F. J.; Morgan, G. L.

    2003-03-01

    Post-processed total neutron, RIF neutron, gamma-ray, and x-ray images from 2D LASNEX calculations of burning ignition capsules are presented. The capsules have yields ranging from tens of kilojoules (failures) to over 16 MJ (ignition), and their implosion symmetry ranges from prolate (flattest at the hohlraum equator) to oblate (flattest towards the laser entrance hole). The simulated total neutron images emphasize regions of high DT density and temperature; the reaction-in-flight neutrons emphasize regions of high DT density; the gamma rays emphasize regions of high shell density; and the high energy x rays (>10 keV) emphasize regions of high temperature.

  6. Measurements of neutron and gamma ray streaming through a duct, (2), (3)

    International Nuclear Information System (INIS)

    Hashikura, Hiroyuki; Fukumoto, Hideshi; Akiyama, Masatsugu; Oka, Yoshiaki; An, Shigehiro

    1982-03-01

    Measurements of neutron and gamma ray streaming through a duct measurements of and a cavity in concrete shields were measured in the fast neutron source reactor YAYOI of the University of Tokyo. The neutron spectra measured by a NE213 scintillator and proton recoil counters were compared with the calculations using Monte Carlo code, MORSE-CG. The agreements between the experiments and the calculations were generally satisfactory. The attenuations of neutron and gamma ray in the cavity and the duct were studied in the three experimental configurations. (author)

  7. Micronuclei induced by fast neutrons versus 60Co gamma-rays in human peripheral blood lymphocytes.

    Science.gov (United States)

    Vral, A; Verhaegen, F; Thierens, H; De Ridder, L

    1994-03-01

    Here we compared the effectiveness of neutrons ( = 5.5 MeV) versus 60Co gamma-rays in producing micronuclei (MN) in human lymphocytes. To obtain dose-response data, blood samples of six donors were irradiated with doses ranging from 0.1 to 5 Gy for gamma-rays and 0.1-3 Gy for neutrons. A linear dependence of MN yield with dose was found for fast neutrons while for gamma-rays a nonlinear dependence existed. For both radiation qualities no significant interindividual differences were found. Derived relative biological effectiveness values decreased with increasing dose. The MN frequency distributions were overdispersed with respect to the Poisson distribution, with neutrons showing higher dispersion values than with gamma-rays. To compare the repair kinetics of both radiation qualities split-dose experiments were performed. A dose of 4 Gy gamma-rays (3 Gy neutrons) was delivered either as a single exposure or in two equal fractions separated by time intervals ranging from 30 min to 10 h (30 min to 7 h for neutrons). The data showed for gamma-rays a significant decline (30% +/- 10%) in MN yield with interfraction time due to repair of DNA damage. This repair is a continuous process starting almost immediately after the first of the two doses and lasting 3-5 h. For fast neutrons no decline was observed indicating irreparable damage.

  8. A passive-active neutron device for assaying remote-handled transuranic waste

    International Nuclear Information System (INIS)

    Estep, R.J.; Coop, K.L.; Deane, T.M.; Lujan, J.E.

    1990-01-01

    A combined passive-active neutron assay device was constructed for assaying remote-handled transuranic waste. A study of matrix and source position effects in active assays showed that a knowledge of the source position alone is not sufficient to correct for position-related errors in highly moderating or absorbing matrices. An alternate function for the active assay of solid fuel pellets was derived, although the efficacy of this approach remains to be established

  9. Evaluation of gamma and neutron irradiation effects on the properties of mica film capacitors

    International Nuclear Information System (INIS)

    Roy, Rajesh; Pandya, Arun

    2005-01-01

    We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 μm (0.7874 and 1.5748 mil) with two different areas, 01 and 04 cm 2 . The capacitance has been measured at room temperature in the frequency range 100 Hz-10 MHz. Negligible change in the capacitance due to high gamma dose of 60 Co, 15 kGy at dose rate 0.25 kGy/h, has been observed. However, appreciable change in the capacitance has been observed due to low doses of fast neutrons (cumulative dose, 115 cGy) with flux ∼ 9.925 X 10 7 neutrons/cm 2 h from 252 Cf neutron source of fluence, 2.5 x 10 7 neutrons/s. We have also observed that the impact of gamma and neutron irradiation is more at frequencies higher than 10 kHz, These results show that the mica capacitors do not show any radiation response below 10 kHz. The study shows the radiation response of mica film capacitors to gamma and fast neutron radiations. Mica capacitors show low gamma radiation response in comparison to fast neutron radiation, because a total dose of kGy order has been given by gamma source and only few cGy dose has been given by fast neutron source. (author)

  10. Use of the johnin PPD interferon-gamma assay in control of bovine paratuberculosis

    DEFF Research Database (Denmark)

    Jungersen, Gregers; Mikkelsen, Heidi; Grell, Susanne N.

    2012-01-01

    Although the interferon-gamma (IFN-γ) assay for measurements of cell-mediated immune (CMI) responses to paratuberculosis PPD (johnin) has been available for close to 20 years, the assay has not yet emerged as the long desired test to identify infected animals at an early time point. Among other...

  11. Time Evolving Fission Chain Theory and Fast Neutron and Gamma-Ray Counting Distributions

    International Nuclear Information System (INIS)

    Kim, K. S.; Nakae, L. F.; Prasad, M. K.; Snyderman, N. J.; Verbeke, J. M.

    2015-01-01

    Here, we solve a simple theoretical model of time evolving fission chains due to Feynman that generalizes and asymptotically approaches the point model theory. The point model theory has been used to analyze thermal neutron counting data. This extension of the theory underlies fast counting data for both neutrons and gamma rays from metal systems. Fast neutron and gamma-ray counting is now possible using liquid scintillator arrays with nanosecond time resolution. For individual fission chains, the differential equations describing three correlated probability distributions are solved: the time-dependent internal neutron population, accumulation of fissions in time, and accumulation of leaked neutrons in time. Explicit analytic formulas are given for correlated moments of the time evolving chain populations. The equations for random time gate fast neutron and gamma-ray counting distributions, due to randomly initiated chains, are presented. Correlated moment equations are given for both random time gate and triggered time gate counting. There are explicit formulas for all correlated moments are given up to triple order, for all combinations of correlated fast neutrons and gamma rays. The nonlinear differential equations for probabilities for time dependent fission chain populations have a remarkably simple Monte Carlo realization. A Monte Carlo code was developed for this theory and is shown to statistically realize the solutions to the fission chain theory probability distributions. Combined with random initiation of chains and detection of external quanta, the Monte Carlo code generates time tagged data for neutron and gamma-ray counting and from these data the counting distributions.

  12. Use of prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent

    Energy Technology Data Exchange (ETDEWEB)

    Priyada, P.; Sarkar, P.K., E-mail: pradip.sarkar@manipal.edu

    2015-06-11

    The possibility of using measured prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent is explored theoretically. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of a high density polyethylene cylinder to emit prompt gammas from interaction of neutrons with the nuclei of hydrogen and carbon present in polyethylene. The neutron energy dependent responses of hydrogen and carbon nuclei are combined appropriately to match the energy dependent neutron fluence to ambient dose equivalent conversion coefficients. The proposed method is tested initially with simulated spectra and then validated using experimental measurements with an Am–Be neutron source. Experimental measurements and theoretical simulations have established the feasibility of estimating neutron ambient dose equivalent using measured neutron induced prompt gammas emitted from polyethylene with an overestimation of neutron dose at very low energies. - Highlights: • A new method for estimating H{sup ⁎}(10) using prompt gamma emissions from HDPE. • Linear combination of 2.2 MeV and 4.4 MeV gamma intensities approximates DCC (ICRP). • Feasibility of the method was established theoretically and experimentally. • The response of the present technique is very similar to that of the rem meters.

  13. Nondestructive assay of TRU waste using gamma-ray active and passive computed tomography

    International Nuclear Information System (INIS)

    Roberson, G.P.; Decman, D.; Martz, H.; Keto, E.R.; Johansson, E.M.

    1995-01-01

    The authors have developed an active and passive computed tomography (A and PCT) scanner for assaying radioactive waste drums. Here they describe the hardware components of their system and the software used for data acquisition, gamma-ray spectroscopy analysis, and image reconstruction. They have measured the performance of the system using ''mock'' waste drums and calibrated radioactive sources. They also describe the results of measurements using this system to assay a real TRU waste drum with relatively low Pu content. The results are compared with X-ray NDE studies of the same TRU waste drum as well as assay results from segmented gamma scanner (SGS) measurements

  14. Next Generation Gamma/Neutron Detectors for Planetary Science., Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Gamma ray and neutron spectroscopy are well established techniques for determining the chemical composition of planetary surfaces, and small cosmic bodies such as...

  15. Resistive plate chamber neutron and gamma sensitivity measurement with a {sup 252}Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Abbrescia, M.; Altieri, S.; Baratti, V.; Barnaba, O.; Belli, G.; Bruno, G.; Colaleo, A.; DeVecchi, C.; Guida, R. E-mail: roberto.guida@pv.infn.it; Iaselli, G.; Imbres, E.; Loddo, F.; Maggi, M.; Marangelli, B.; Musitelli, G.; Nardo, R.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Ratti, S.; Riccardi, C.; Romano, F.; Torre, P.; Vicini, A.; Vitulo, P.; Volpe, F

    2003-06-21

    A bakelite double gap Resistive Plate Chamber (RPC), operating in avalanche mode, has been exposed to the radiation emitted from a {sup 252}Cf source to measure its neutron and gamma sensitivity. One of the two gaps underwent the traditional electrodes surface coating with linseed oil. RPC signals were triggered by fission events detected using BaF{sub 2} scintillators. A Monte Carlo code, inside the GEANT 3.21 framework with MICAP interface, has been used to identify the gamma and neutron contributions to the total number of collected RPC signals. A neutron sensitivity of (0.63{+-}0.02)x10{sup -3} (average energy 2 MeV) and a gamma sensitivity of (14.0{+-}0.5)x10{sup -3} (average energy 1.5 MeV) have been measured in double gap mode. Measurements done in single gap mode have shown that both neutron and gamma sensitivity are independent of the oiling treatment.

  16. Neutron-gamma discrimination based on pulse shape discrimination in a Ce:LiCaAlF{sub 6} scintillator

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Atsushi, E-mail: a-yamazaki@nucl.nagoya-u.ac.jp [Department of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University (Japan); Watanabe, Kenichi; Uritani, Akira [Department of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University (Japan); Iguchi, Tetsuo [Department of Quantum Engineering, Graduate School of Engineering, Nagoya University (Japan); Kawaguchi, Noriaki [Tokuyama Corporation (Japan); Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei [Institute of Multidisciplinary Research for Advanced Materials (IMRAM), Tohoku University (Japan); Fukuda, Kentaro; Suyama, Toshihisa [Tokuyama Corporation (Japan); Yoshikawa, Akira [Institute of Multidisciplinary Research for Advanced Materials (IMRAM), Tohoku University (Japan); New Industry Creation Hatchery Center (NICHe), Tohoku University (Japan)

    2011-10-01

    We demonstrate neutron-gamma discrimination based on a pulse shape discrimination method in a Ce:LiCAF scintillator. We have tried neutron-gamma discrimination using a difference in the pulse shape or the decay time of the scintillation light pulse. The decay time is converted into the rise time through an integrating circuit. A {sup 252}Cf enclosed in a polyethylene container is used as the source of thermal neutrons and prompt gamma-rays. Obvious separation of neutron and gamma-ray events is achieved using the information of the rise time of the scintillation light pulse. In the separated neutron spectrum, the gamma-ray events are effectively suppressed with little loss of neutron events. The pulse shape discrimination is confirmed to be useful to detect neutrons with the Ce:LiCAF scintillator under an intense high-energy gamma-ray condition.

  17. Unilateral irradiation of pigs in a mixed neutrons+gamma field. Early results

    International Nuclear Information System (INIS)

    Lemaitre, Guy; Maas, Jean.

    1982-08-01

    Pigs (16-20kg) were irradiated with 60 Co gamma or in a mixed field (neutron + gamma from the pulsed reactor SILENE). Pigs were unilaterally exposed by the left side. Each experimental group was composed of twelve animals and one control. Within the dose range explored (reference dose is mid-line tissue dose): 4-9.8 Gy of gamma rays only; 4.6 - 5.7 Gy of neutrons and gamma rays, pigs presented the haematopioetic form of the acute radiation sickness. At 5 Gy mixed field was more harmful than gamma rays only. Therefore the numerical value of neutron RBE (lethality 50 p cent within 30 days) is more than one. Experiments will be carried out in order to determine RBE values more accurately. Bone marrow dose will also be determined [fr

  18. Study of associated gamma from niobium under 14. 9 MeV neutron bombardments

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Hongyu; Yan Yiming; Fan Guoying; Lan Liqiac; Sun Suxu; Wang Qi; Hua Ming; Han Chongzhen; Liu Shuzhenn; Rong Yaning; and others

    1989-02-01

    The gamma ray spectra from niobium under 14.9 MeV neutron bombardments were measured by means of a pulsed /ital T/(/ital d/, /ital n/)/sup 4/He neutron source, associated particle method, Ge(Li) detector and time-of-flight technique at 7 angles between 30/degree/ and 140/degree/. 79 gamma lines were determined by a high resolution gamma spectrum analysis program, and reaction types and transition levels of 62 lines were roughly assigned. There were 40 ones of 79 lines, which were first found in reactions induced by neutrons. The differential cross sections of every gamma line at 7 angles were determined. It is shown that associated gamma ray emissions from this reaction are basically isotropic.

  19. Tangential channel for nuclear gamma-resonance spectroscopy in thermal neutron capture

    International Nuclear Information System (INIS)

    Belogurov, V.N.; Bondars, H.Ya.; Lapenas, A.A.; Reznikov, R.S.; Senkov, P.E.

    1979-01-01

    Design of a tangential reactor channel which has been made to replace the radial one in the pulsed research reactor IRT-2000 is described. It allows to use the same hole in biological reactor schielding. Characteristics of neutron and gamma-background spectra at the excit of the channel are given and compared with analogous characteristics of the radial one. The gamma background in the tangential channel is lower than in the radial channel. The gamma spectra in the Gd 155 (n, γ)Gd 156 , Gd 157 (n, γ)Gd 158 , Er 167 (n, γ)Er 168 and Hf 177 (n, γ)Hf 178 reactions show that the application of X-ray detection units BDR with the tangential channel allows to carry out the gamma spectrometry of gamma quanta emitted in the thermal neutron capture by both high and low neutron capture cross section nuclei (e.g., Gdsup(157, 155) and Er 167 , Hf 177 , respectively)

  20. Measurements of the gamma-quanta angular distributions emitted from neutron inelastic scattering on 28Si

    Science.gov (United States)

    Fedorov, N. A.; Grozdanov, D. N.; Bystritskiy, V. M.; Kopach, Yu. N.; Ruskov, I. N.; Skoy, V. R.; Tretyakova, T. Yu.; Zamyatin, N. I.; Wang, D.; Aliev, F. A.; Hramco, C.; Gandhi, A.; Kumar, A.; Dabylova, S.; Bogolubov, E. P.; Barmakov, Yu. N.

    2018-04-01

    The characteristic gamma radiation from the interaction of 14.1 MeV neutrons with a natural silicon sample is investigated with Tagged Neutron Method (TNM). The anisotropy of gamma-ray emission of 1.779 MeV was measured at 11 azimuth angles with a step of ∠15°. The present results are in good agreement with some recent experimental data.

  1. Energy–angle correlation of neutrons and gamma-rays emitted from an HEU source

    Energy Technology Data Exchange (ETDEWEB)

    Miloshevsky, G., E-mail: gennady@purdue.edu; Hassanein, A.

    2014-06-01

    Special Nuclear Materials (SNM) yield very unique fission signatures, namely correlated neutrons and gamma-rays. A major challenge is not only to detect, but also to rapidly identify and recognize SNM with certainty. Accounting for particle multiplicity and correlations is one of standard ways to detect SNM. However, many parameter data such as joint distributions of energy, angle, lifetime, and multiplicity of neutrons and gamma-rays can lead to better recognition of SNM signatures in the background radiation noise. These joint distributions are not well understood. The Monte Carlo simulations of the transport of neutrons and gamma-rays produced from spontaneous and interrogation-induced fission of SNM are carried out using the developed MONSOL computer code. The energy spectra of neutrons and gamma-rays from a bare Highly Enriched Uranium (HEU) source are investigated. The energy spectrum of gamma-rays shows spectral lines by which HEU isotopes can be identified, while those of neutrons do not show any characteristic lines. The joint probability density function (JPDF) of the energy–angle association of neutrons and gamma-rays is constructed. Marginal probability density functions (MPDFs) of energy and angle are derived from JPDF. A probabilistic model is developed for the analysis of JPDF and MPDFs. This probabilistic model is used to evaluate mean values, standard deviations, covariance and correlation between the energy and angle of neutrons and gamma-rays emitted from the HEU source. For both neutrons and gamma-rays, it is found that the energy–angle variables are only weakly correlated.

  2. Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

    International Nuclear Information System (INIS)

    Lefvert, T.

    1975-11-01

    Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

  3. Creation and validation of a neutron-gamma coupled multigroup cross section library

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Lee, S.M.

    1995-01-01

    The task of creating our own neutron-gamma coupled library was taken up. By using 1985 version of NJOY code system, a coupled set called IGC-DE4-S1 in ANISN format for 25 nuclides has been arrived at based on ENDF/B-IV neutron library and DLC-99 gamma library, with Legendre order of up to 5. The flow chart for the creation of coupled set is given. 5 refs, 1 fig., 3 tabs

  4. Measurement of neutron and gamma-ray production double differential cross section at KEK

    International Nuclear Information System (INIS)

    Ishibashi, Kenji

    1995-01-01

    High energy nuclear radiations were measured for 0.8-3.0 GeV proton induced reactions at KEK. The measurement was carried out to overcome the problems arising from the use of secondary beam line of a quite low incident beam intensity. Digital pulse shape discrimination method was applicable to separation between high energy neutrons and gamma-rays. By the use of a number of scintillators, cross sections were obtained for production of neutrons and gamma-rays. (author)

  5. Control of radioactive wastes and coupling of neutron/gamma measurements: use of radiative capture for the correction of matrix effects that penalize the fissile mass measurement by active neutron interrogation; Controle des dechets radioactifs et couplage de mesures neutron/gamma: exploitation de la capture radiative pour corriger les effets de matrice penalisant la mesure de la masse fissile par interrogation neutronique active

    Energy Technology Data Exchange (ETDEWEB)

    Loche, F

    2006-10-15

    In the framework of radioactive waste drums control, difficulties arise in the nondestructive measurement of fissile mass ({sup 235}U, {sup 239}Pu..) by Active Neutron Interrogation (ANI), when dealing with matrices containing materials (Cl, H...) influencing the neutron flux. The idea is to use the neutron capture reaction (n,{gamma}) to determine the matrix composition to adjust the ANI calibration coefficient value. This study, dealing with 118 litres, homogeneous drums of density less than 0,4 and composed of chlorinated and/or hydrogenated materials, leads to build abacus linking the {gamma} ray peak areas to the ANI calibration coefficient. Validation assays of these abacus show a very good agreement between the corrected and true fissile masses for hydrogenated matrices (max. relative standard deviation: 23 %) and quite good for chlorinated and hydrogenated matrices (58 %). The developed correction method improves the measured values. It may be extended to 0,45 density, heterogeneous drums. (author)

  6. Detection of gamma-neutron radiation by solid-state scintillation detectors. Detection of gamma-neutron radiation by novel solid-state scintillation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Ryzhikov, V.; Grinyov, B.; Piven, L.; Onyshchenko, G.; Sidletskiy, O. [Institute for Scintillation Materials of the NAS of Ukraine, Kharkov, (Ukraine); Naydenov, S. [Institute for Single Crystals of the National Academy of Sciences of Ukraine, Kharkov, (Ukraine); Pochet, T. [DETEC-Europe, Vannes (France); Smith, C. [Naval Postgraduate School, Monterey, CA (United States)

    2015-07-01

    It is known that solid-state scintillators can be used for detection of both gamma radiation and neutron flux. In the past, neutron detection efficiencies of such solid-state scintillators did not exceed 5-7%. At the same time it is known that the detection efficiency of the gamma-neutron radiation characteristic of nuclear fissionable materials is by an order of magnitude higher than the efficiency of detection of neutron fluxes alone. Thus, an important objective is the creation of detection systems that are both highly efficient in gamma-neutron detection and also capable of exhibiting high gamma suppression for use in the role of detection of neutron radiation. In this work, we present the results of our experimental and theoretical studies on the detection efficiency of fast neutrons from a {sup 239}Pu-Be source by the heavy oxide scintillators BGO, GSO, CWO and ZWO, as well as ZnSe(Te, O). The most probable mechanism of fast neutron interaction with nuclei of heavy oxide scintillators is the inelastic scattering (n, n'γ) reaction. In our work, fast neutron detection efficiencies were determined by the method of internal counting of gamma-quanta that emerge in the scintillator from (n, n''γ) reactions on scintillator nuclei with the resulting gamma energies of ∼20-300 keV. The measured efficiency of neutron detection for the scintillation crystals we considered was ∼40-50 %. The present work included a detailed analysis of detection efficiency as a function of detector and area of the working surface, as well as a search for new ways to create larger-sized detectors of lower cost. As a result of our studies, we have found an unusual dependence of fast neutron detection efficiency upon thickness of the oxide scintillators. An explanation for this anomaly may involve the competition of two factors that accompany inelastic scattering on the heavy atomic nuclei. The transformation of the energy spectrum of neutrons involved in the (n, n

  7. A l-nCi/g sensitivity transuranic waste assay system using pulsed neutron interrogation

    International Nuclear Information System (INIS)

    Kunz, W.E.; Atencio, J.D.; Caldwell, J.T.

    1980-01-01

    We have developed a pulsed thermal neutron interrogation system and have demonstrated a sub-1-nCi/g assay sensitivity for high density TRU wastes contained in 200-liter barrels. We detect prompt fission neutrons, resulting in greatly enhanced sensitivity compared to techniques in which delayed fission neutrons are detected. We observe a linear assay response over at least three orders of magnitude in 235 U (or 239 Pu) mass. We also have measured a flat (to +-10%) interrogation flux profile throughout the volume of a 200-liter barrel filled with 200 kg of sand and vermiculite, which indicates flatness of response to fissile material at different locations within the barrel

  8. Analysis of unstable chromosome alterations frequency induced by neutron-gamma mixed field radiation

    International Nuclear Information System (INIS)

    Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Vale, Carlos H.F.P.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

    2009-01-01

    Nowadays monitoring chromosome alterations in peripheral blood lymphocytes have been used to access the radiation absorbed dose in individuals exposed accidental or occupationally to gamma radiation. However there are not many studies based on the effects of mixed field neutron-gamma. The radiobiology of neutrons has great importance because in nuclear factories worldwide there are several hundred thousand individuals monitored as potentially receiving doses of neutron. In this paper it was observed the frequencies of unstable chromosome alterations induced by a gamma-neutron mixed field. Blood was obtained from one healthy donor and exposed to mixed field neutron-gamma sources 241 AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphases were analyzed for the presence of chromosome alterations by two experienced scorers. The results suggest that there is the possibility of a directly proportional relationship between absorbed dose of neutron-gamma mixed field radiation and the frequency of unstable chromosome alterations analyzed in this paper. (author)

  9. Analysis of unstable chromosome alterations frequency induced by neutron-gamma mixed field radiation

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Priscilla L.G.; Brandao, Jose Odinilson de C.; Vale, Carlos H.F.P.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)], e-mail: psouza@cnen.gov.br, e-mail: jodinilson@cnen.gov.br; Calixto, Merilane S.; Santos, Neide [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Genetica

    2009-07-01

    Nowadays monitoring chromosome alterations in peripheral blood lymphocytes have been used to access the radiation absorbed dose in individuals exposed accidental or occupationally to gamma radiation. However there are not many studies based on the effects of mixed field neutron-gamma. The radiobiology of neutrons has great importance because in nuclear factories worldwide there are several hundred thousand individuals monitored as potentially receiving doses of neutron. In this paper it was observed the frequencies of unstable chromosome alterations induced by a gamma-neutron mixed field. Blood was obtained from one healthy donor and exposed to mixed field neutron-gamma sources {sup 241}AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphases were analyzed for the presence of chromosome alterations by two experienced scorers. The results suggest that there is the possibility of a directly proportional relationship between absorbed dose of neutron-gamma mixed field radiation and the frequency of unstable chromosome alterations analyzed in this paper. (author)

  10. Self-Powered Neutron and Gamma Detectors for In-Core Measurements

    International Nuclear Information System (INIS)

    Strindehag, O.

    1971-11-01

    The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

  11. Self-Powered Neutron and Gamma Detectors for In-Core Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1971-11-15

    The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

  12. PANDORA, a large volume low-energy neutron detector with real-time neutron-gamma discrimination

    Science.gov (United States)

    Stuhl, L.; Sasano, M.; Yako, K.; Yasuda, J.; Baba, H.; Ota, S.; Uesaka, T.

    2017-09-01

    The PANDORA (Particle Analyzer Neutron Detector Of Real-time Acquisition) system, which was developed for use in inverse kinematics experiments with unstable isotope beams, is a neutron detector based on a plastic scintillator coupled to a digital readout. PANDORA can be used for any reaction study involving the emission of low energy neutrons (100 keV-10 MeV) where background suppression and an increased signal-to-noise ratio are crucial. The digital readout system provides an opportunity for pulse shape discrimination (PSD) of the detected particles as well as intelligent triggering based on PSD. The figure of merit results of PANDORA are compared to the data in literature. Using PANDORA, 91 ± 1% of all detected neutrons can be separated, while 91 ± 1% of the detected gamma rays can be excluded, reducing the gamma ray background by one order of magnitude.

  13. Stability evaluation and correction of a pulsed neutron generator prompt gamma activation analysis system

    Science.gov (United States)

    Source output stability is important for accurate measurement in prompt gamma neutron activation. This is especially true when measuring low-concentration elements such as in vivo nitrogen (~2.5% of body weight). We evaluated the stability of the compact DT neutron generator within an in vivo nitrog...

  14. Neutron activation analysis of lipsticks using gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Mirsa, G.; Mittal, V.K.

    2004-01-01

    Neutron activation analysis with gamma-ray spectrometry was used to measure the concentrations of various elements in lipsticks of popular Indian and foreign brands. The aim of the present work was to study the possibility of existence of trace elements in samples of lipsticks (the ingredients of which are mostly organic in nature) and to see whether trace elements could distinguish lipsticks of different Indian and foreign brands from the forensic point of view apart from their inter-se differentiation. In the different samples of lipsticks that were analysed the following elements were detected: Au, Ba, Br, Ca, Cs, Fe, Na, Ru, Sb, Sc, Ta, Yb, Zn, Rb and Se. It was found that inter-se differentiation of lipsticks was possible on the basis of concentrations of trace elements and their profile. Concentration of bromine in samples of lipsticks identified lipsticks of different Indian brands. Samples of lipsticks of Indian and foreign brands could be differentiated on the basis of concentrations of cesium, antimony and scandium which were found to be higher in foreign brands as compared to those in Indian brands. (authors)

  15. Automated gamma spectrometry and data analysis on radiometric neutron dosimeters

    International Nuclear Information System (INIS)

    Matsumoto, W.Y.

    1983-01-01

    An automated gamma-ray spectrometry system was designed and implemented by the Westinghouse Hanford Company at the Hanford Engineering Development Laboratory (HEDL) to analyze radiometric neutron dosimeters. Unattended, automatic, 24 hour/day, 7 day/week operation with online data analysis and mainframe-computer compatible magnetic tape output are system features. The system was used to analyze most of the 4000-plus radiometric monitors (RM's) from extensive reactor characterization tests during startup and initial operation of th Fast Flux Test Facility (FFTF). The FFTF, operated by HEDL for the Department of Energy, incorporates a 400 MW(th) sodium-cooled fast reactor. Aumomated system hardware consists of a high purity germanium detector, a computerized multichannel analyzer data acquisition system (Nuclear Data, Inc. Model 6620) with two dual 2.5 Mbyte magnetic disk drives plus two 10.5 inch reel magnetic tape units for mass storage of programs/data and an automated Sample Changer-Positioner (ASC-P) run with a programmable controller. The ASC-P has a 200 sample capacity and 12 calibrated counting (analysis) positions ranging from 6 inches (15 cm) to more than 20 feet (6.1 m) from the detector. The system software was programmed in Fortran at HEDL, except for the Nuclear Data, Inc. Peak Search and Analysis Program and Disk Operating System (MIDAS+)

  16. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    J.S. Tang

    2001-05-03

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

  17. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    International Nuclear Information System (INIS)

    J.S. Tang

    2001-01-01

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

  18. Classification of JET Neutron and Gamma Emissivity Profiles

    Science.gov (United States)

    Craciunescu, T.; Murari, A.; Kiptily, V.; Vega, J.; Contributors, JET

    2016-05-01

    In thermonuclear plasmas, emission tomography uses integrated measurements along lines of sight (LOS) to determine the two-dimensional (2-D) spatial distribution of the volume emission intensity. Due to the availability of only a limited number views and to the coarse sampling of the LOS, the tomographic inversion is a limited data set problem. Several techniques have been developed for tomographic reconstruction of the 2-D gamma and neutron emissivity on JET. In specific experimental conditions the availability of LOSs is restricted to a single view. In this case an explicit reconstruction of the emissivity profile is no longer possible. However, machine learning classification methods can be used in order to derive the type of the distribution. In the present approach the classification is developed using the theory of belief functions which provide the support to fuse the results of independent clustering and supervised classification. The method allows to represent the uncertainty of the results provided by different independent techniques, to combine them and to manage possible conflicts.

  19. Gamma spectrum following neutron capture in {sup 167}Er

    Energy Technology Data Exchange (ETDEWEB)

    Visser, D.; Khoo, T.L.; Lister, C.J. [and others

    1995-08-01

    Statistical decay from a highly excited state samples all the lower-lying states and, hence, provides a sensitive measure of the level density. Pairing has a major impact on the level density, e.g. creating a pair gap between the 0- and 2-quasiparticle configurations. Hence the shape of the statistical spectrum contains information on pairing, and can be used to provide information on the reduction of pairing with thermal excitation energy. For this reason, we measured the complete spectrum of {gamma}rays following thermal neutron capture in {sup 167}Er. The experiment was performed at the Brookhaven reactor using Compton-suppressed Ge detectors from TESSA. The spectrum, which was corrected for detector response and efficiency, reveals primary (first-step, high-energy) transitions up to nearly 8 MeV, secondary (last-step, lower-energy) transitions, as we as a continuous statistical component. Effort was expanded to identify all lines from contaminant sources and an upper limit of 5% was tentatively set for their contributions. The spectral shape of the statistical spectrum will be compared with theoretical spectra obtained from a calculation of pairing which accounts for a stepwise reduction of the pair correlations as the number of quasiparticles increases. The primary lines which decay directly to the near-yrast states will also be used to deduce the level densities.

  20. Experimental arrangement for production and use of gamma radiation from neutron capture

    International Nuclear Information System (INIS)

    Mafra, Olga Yajgunovitch

    1969-01-01

    This dissertation presents the main characteristics and construction details of collimator system for gamma radiation emitted by atomic nuclei after capturing thermal neutrons. This construction was made in one of the cross channels of IEAR-1 swimming pool reactor of the Atomic Energy Institute of Sao Paulo, Brazil. The energies of gamma radiation available vary range from about 4 MeV and 11 MeV, discreetly. With this experimental arrangement is obtained: high intensity, good collimation and monochrome gamma radiation, important for conducting experiments with gamma radiation. It is also present in this dissertation the description of the techniques employed in determining the intensity of gamma radiation and the extent of contamination in the neutron beam as well as the program list GAMAU that adjusts the gamma spectrum photopeak taken as a Gaussian curve. We intend to use this experimental arrangement for the measurement of cross sections of photonuclear reactions

  1. Design of a facility by neutron activation by spectrometry of prompt gamma

    International Nuclear Information System (INIS)

    Oliver, R.; Benites L, S.; Montoya Z, M.

    1993-01-01

    We show the basic design of the facility of PGNAA that we will install in the hall of the peruvian reactor RP-10. The thermal neutron flux (without a gamma filter) will be 2,0 x 10 8 n/cm -2 s -1 at 10 MW of power. The ratio of gamma exposition without gamma filter will be 29 kR/h. (authors). 8 refs., 2 figs

  2. Development of a new deuterium-deuterium (D-D) neutron generator for prompt gamma-ray neutron activation analysis.

    Science.gov (United States)

    Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

    2014-12-01

    A new deuterium-deuterium (D-D) neutron generator has been developed by Adelphi Technology for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA), and fast neutron radiography. The generator makes an excellent fast, intermediate, and thermal neutron source for laboratories and industrial applications that require the safe production of neutrons, a small footprint, low cost, and small regulatory burden. The generator has three major components: a Radio Frequency Induction Ion Source, a Secondary Electron Shroud, and a Diode Accelerator Structure and Target. Monoenergetic neutrons (2.5MeV) are produced with a yield of 10(10)n/s using 25-50mA of deuterium ion beam current and 125kV of acceleration voltage. The present study characterizes the performance of the neutron generator with respect to neutron yield, neutron production efficiency, and the ionic current as a function of the acceleration voltage at various RF powers. In addition the Monte Carlo N-Particle Transport (MCNP) simulation code was used to optimize the setup with respect to thermal flux and radiation protection. Copyright © 2014 Elsevier Ltd. All rights reserved.

  3. Development of a lion-specific interferon-gamma assay

    NARCIS (Netherlands)

    Maas, M.; Kooten, van P.J.S.; Schreuder, J.; Morar, D.; Tijhaar, E.; Michel, A.L.; Rutten, V.P.M.G.

    2012-01-01

    The ongoing spread of bovine tuberculosis (BTB) in African free-ranging lion populations, for example in the Kruger National Park, raises the need for diagnostic assays for BTB in lions. These, in addition, would be highly relevant for zoological gardens worldwide that want to determine the BTB

  4. Active neutron and gamma-ray imaging of highly enriched uranium for treaty verification.

    Science.gov (United States)

    Hamel, Michael C; Polack, J Kyle; Ruch, Marc L; Marcath, Matthew J; Clarke, Shaun D; Pozzi, Sara A

    2017-08-11

    The detection and characterization of highly enriched uranium (HEU) presents a large challenge in the non-proliferation field. HEU has a low neutron emission rate and most gamma rays are low energy and easily shielded. To address this challenge, an instrument known as the dual-particle imager (DPI) was used with a portable deuterium-tritium (DT) neutron generator to detect neutrons and gamma rays from induced fission in HEU. We evaluated system response using a 13.7-kg HEU sphere in several configurations with no moderation, high-density polyethylene (HDPE) moderation, and tungsten moderation. A hollow tungsten sphere was interrogated to evaluate the response to a possible hoax item. First, localization capabilities were demonstrated by reconstructing neutron and gamma-ray images. Once localized, additional properties such as fast neutron energy spectra and time-dependent neutron count rates were attributed to the items. For the interrogated configurations containing HEU, the reconstructed neutron spectra resembled Watt spectra, which gave confidence that the interrogated items were undergoing induced fission. The time-dependent neutron count rate was also compared for each configuration and shown to be dependent on the neutron multiplication of the item. This result showed that the DPI is a viable tool for localizing and confirming fissile mass and multiplication.

  5. The efficient neutron-gamma pulse shape discrimination with small active volume scintillation detector

    International Nuclear Information System (INIS)

    Phan Van Chuan; Nguyen Duc Hoa; Nguyen Xuan Hai; Nguyen Ngoc Anh; Tuong Thi Thu Huong; Nguyen Nhi Dien; Pham Dinh Khang

    2016-01-01

    A small detector with EJ-301 liquid scintillation was manufactured for the study on the neutron-gamma pulse shape discrimination. In this research, four algorithms, including Threshold crossing time (TCT), Pulse gradient analysis (PGA), Charge comparison method (CCM), and Correlation pattern recognition (CPR) were developed and compared in terms of their discrimination effectiveness between neutrons and gamma rays. The figures of merits (FOMs) obtained for 100 ÷ 2000 keVee (keV energy electron equivalent) neutron energy range show the charge comparison method was the most efficient of the four algorithms. (author)

  6. A gamma/neutron-discriminating, Cooled, Optically Stimulated Luminescence (COSL) dosemeter

    International Nuclear Information System (INIS)

    Eschbach, P.A.; Miller, S.D.

    1992-07-01

    The Cooled Optically Stimulated Luminescence (COSL) of CaF 2 :Mn (grain sizes from 0.1 to 100 microns) powder embedded in a hydrogenous matrix is reported as a function of fast-neutron dose. When all the CaF 2 :Mn grains are interrogated at once, the COSL plastic dosemeters have a minimum detectable limit of 1 cSv fast neutrons; the gamma component from the bare 252 cf exposure was determined with a separate dosemeter. We report here on a proton-recoil-based dosemeter that generates pulse height spectra, much like the scintillator of Hornyak, (2) to provide information on both the neutron and gamma dose

  7. Baseline distortion effect on gamma-ray pulse-height spectra in neutron capture experiments

    International Nuclear Information System (INIS)

    Laptev, A.; Harada, H.; Nakamura, S.; Hori, J.; Igashira, M.; Ohsaki, T.; Ohgama, K.

    2005-01-01

    A baseline distortion effect due to gamma-flash at neutron time-of-flight measurement using a pulse neutron source has been investigated. Pulses from C 6 D 6 detectors accumulated by flash-ADC were processed with both standard analog-to-digital converter (ADC) and flash-ADC operational modes. A correction factor of gamma-ray yields, due to baseline shift, was quantitatively obtained by comparing the pulse height spectra of the two data-taking modes. The magnitude of the correction factor depends on the time after gamma-flash and has complex time dependence with a changing sign

  8. A FIFO based neutron arrival time collection technique for assay of plutonium

    International Nuclear Information System (INIS)

    Parthasarathy, R.; Saisubalakshmi, D.; Venkatasubramani, C.R.

    2004-01-01

    The system assays plutonium by counting the time correlated neutrons emitted by the spontaneous fissions of the even-even Pu isotopes in the presence of random neutron background, originating principally from (a,n) reactions in the material. The correlation technique discussed in this paper utilizes twofold neutron coincidence counting but the system is proposed to be enhanced for neutron multiplicity counting. A microcontroller based data acquisition system has been developed using a couple of fast FIFO 2kX9 bit memory ICs and a 16 bit counter for identifying time-correlated neutrons. Since the neutron pulses are arriving at a rapid rate, the incoming pulses are buffered in the FIFO and then transferred to PC by the microcontroller through the parallel port. The correlation analysis based on this time arrival information is done in the PC off-line. (author)

  9. Analysis of the propagation of neutrons and gamma-rays from the fast neutron source reactor YAYOI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Shigeo, E-mail: neutron@keyaki.cc.u-tokai.ac.jp [Department of Energy Science and Engineering, School of Engineering, Tokai University, Hiratsuka, Kanagawa 259-1292 (Japan); Murata, Isao [Division of Electrical, Electronic and Information Engineering, Osaka University, Suita, Osaka 565-0871 (Japan); Nakagawa, Tsutomu; Saito, Isao [Nuclear Professional School, School of Engineering, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki 319-1188 (Japan)

    2011-10-01

    The skyshine effect is crucial for designing appropriate shielding. To investigate the skyshine effect, the propagation of neutrons was measured and analyzed at the fast neutron source reactor YAYOI. Pulse height spectra and dose distributions of neutron and secondary gamma-ray were measured outside YAYOI, and analyzed with MCNP-5 and JENDL-3.3. Comparison with the experimental results showed good agreement. Also, a semi-empirical formula was successfully derived to describe the dose distribution. The formulae can be used to predict the skyshine effect at YAYOI, and will be useful for estimating the skyshine effect and designing the shield structure for fusion facilities.

  10. Development of the neutron filters for JET gamma-ray cameras

    International Nuclear Information System (INIS)

    Soare, S.; Curuia, M.; Anghel, M.; Constantin, M.; David, E.; Kiptily, V.; Prior, P.; Edlington, T.; Griph, S.; Krivchenkov, Y.; Popovichev, S.; Riccardo, V.; Syme, B; Thompson, V.; Murari, A.; Zoita, V.; Bonheure, G.; Le Guern

    2007-01-01

    The JET gamma-ray camera diagnostics have already provided valuable information on the gamma-ray imaging of fast ion evaluation in JET plasmas. The JET Gamma-Ray Cameras (GRC) upgrade project deals with the design of appropriate neutron/gamma-ray filters ('neutron attenuaters').The main design parameter was the neutron attenuation factor. The two design solutions, that have been finally chosen and developed at the level of scheme design, consist of: a) one quasi-crescent shaped neutron attenuator (for the horizontal camera) and b) two quasi-trapezoid shaped neutron attenuators (for the vertical one). Various neutron-attenuating materials have been considered (lithium hydride with natural isotopic composition and 6 Li enriched, light and heavy water, polyethylene). Pure light water was finally chosen as the attenuating material for the JET gamma-ray cameras. FEA methods used to evaluate the behaviour of the filter casings under the loadings (internal hydrostatic pressure, torques) have proven the stability of the structure. (authors)

  11. Neutron-induced gamma-ray spectroscopy: simulations for chemical mapping of planetary surfaces

    International Nuclear Information System (INIS)

    Brueckner, J.; Waenke, H.; Reedy, R.C.

    1986-01-01

    Cosmic rays interact with the surface of a planetary body and produce a cascade of secondary particles, such as neutrons. Neutron-induced scattering and capture reactions play an important role in the production of discrete gamma-ray lines that can be measured by a gamma-ray spectrometer on board of an orbiting spacecraft. These data can be used to determine the concentration of many elements in the surface of a planetary body, which provides clues to its bulk composition and in turn to its origin and evolution. To investigate the gamma rays made by neutron interactions, thin targets were irradiated with neutrons having energies from 14 MeV to 0.025 eV. By means of foil activation technique the ratio of epithermal to thermal neutrons was determined to be similar to that in the Moon. Gamma rays emitted by the targets and the surrounding material were detected by a high-resolution germanium detector in the energy range of 0.1 to 8 MeV. Most of the gamma-ray lines that are expected to be used for planetary gamma-ray spectroscopy were found in the recorded spectra and the principal lines in these spectra are presented. 58 refs., 7 figs., 9 tabs

  12. Neutron and gamma sensitivities of self-powered detectors: Monte Carlo modelling

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, Ludo [SCK-CEN, Nuclear Research Centre, Boeretang 200, B-2400 Mol, (Belgium)

    2015-07-01

    This paper deals with the development of a detailed Monte Carlo approach for the calculation of the absolute neutron sensitivity of SPNDs, which makes use of the MCNP code. We will explain the calculation approach, including the activation and beta emission steps, the gamma-electron interactions, the charge deposition in the various detector parts and the effect of the space charge field in the insulator. The model can also be applied for the calculation of the gamma sensitivity of self-powered detectors and for the radiation-induced currents in signal cables. The model yields detailed information on the various contributions to the sensor currents, with distinct response times. Results for the neutron sensitivity of various types of SPNDs are in excellent agreement with experimental data obtained at the BR2 research reactor. For typical neutron to gamma flux ratios, the calculated gamma induced SPND currents are significantly lower than the neutron induced currents. The gamma sensitivity depends very strongly upon the immediate detector surroundings and on the gamma spectrum. Our calculation method opens the way to a reliable on-line determination of the absolute in-pile thermal neutron flux. (authors)

  13. Use of pulsed neutron-neutron logging, thermal neutron-neutron logging, and gamma logging methods in classification for sand-clay sediments of Lower Cretaceous in Prikumsk oil-and-gas region according to filtration-capacitance characteristics

    International Nuclear Information System (INIS)

    Maksimenko, A.N.; Basin, Ya.N.; Novgorodov, V.A.

    1974-01-01

    To isolate reservoirs, the formation and deformation penetration zone parameters are used. They are estimated according to the false oil saturation factor and the time of the penetration zone deformation which are determined from the complex exploration of cased wells using the pulse neutron logging, thermal neutron-neutron logging and gamma logging techniques

  14. Study of gamma ray multiplicity spectra for radiative capture of neutrons in 113,115In

    International Nuclear Information System (INIS)

    Georgiev, G.P.; Fajkov-Stanchik, Kh.; Grigor'ev, Yu.V.; Muradyan, G.V.; Yaneva, N.B.

    1997-08-01

    Neutron radiative capture measurements were performed for the enriched isotopes 113 In and 115 In on the neutron spectrometer at the Neutron Physics Laboratory of the Joint Institute for Nuclear Research employing the gamma ray multiplicity technique and using a ''Romashka'' multi-sectional 4p detector on the 500 m time base of the IBR-30 booster. The gamma multiplicity spectra of resolved resonances were obtained for the 20-500 eV energy range. The mean gamma ray multiplicity was determined for each resonance. The dependence of the ratio S of the low-energy coincidence multiplicity spectrum to the high-energy coincidence multiplicity spectrum on resonance energy exhibits a non-statistical structure. This structure was found to correlate with the local neutron strength function. (author). 10 refs, 6 figs, 2 tabs

  15. Neutron-induced 2.2 MeV background in gamma ray telescopes

    International Nuclear Information System (INIS)

    Zanrosso, E.M.; Long, J.L.; Zych, A.D.; White, R.S.; Hughes Aircraft Co., Los Angeles, CA)

    1985-01-01

    Neutron-induced gamma ray production is an important source of background in Compton scatter gamma ray telescopes where organic scintillator material is used. Most important is deuteron formation when atmospheric albedo and locally produced neutrons are thermalized and subsequently absorbed in the hydrogenous material. The resulting 2.2 MeV gamma line essentially represents a continuous isotropic source within the scintillator itself. Interestingly, using a scintillator material with a high hydrogen-to-carbon ratio to minimize the neutron-induced 4.4 MeV carbon line favors the np reaction. The full problem of neutron-induced background in Compton scatter telescopes has been previously discussed. Results are presented of observations with the University of California balloon-borne Compton scatter telescope where the 2.2 MeV induced line emission is prominently seen

  16. Electrical characterization of commercial NPN bipolar junction transistors under neutron and gamma irradiation

    Directory of Open Access Journals (Sweden)

    OO Myo Min

    2014-01-01

    Full Text Available Electronics components such as bipolar junction transistors, diodes, etc. which are used in deep space mission are required to be tolerant to extensive exposure to energetic neutrons and ionizing radiation. This paper examines neutron radiation with pneumatic transfer system of TRIGA Mark-II reactor at the Malaysian Nuclear Agency. The effects of the gamma radiation from Co-60 on silicon NPN bipolar junction transistors is also be examined. Analyses on irradiated transistors were performed in terms of the electrical characteristics such as current gain, collector current and base current. Experimental results showed that the current gain on the devices degraded significantly after neutron and gamma radiations. Neutron radiation can cause displacement damage in the bulk layer of the transistor structure and gamma radiation can induce ionizing damage in the oxide layer of emitter-base depletion layer. The current gain degradation is believed to be governed by the increasing recombination current in the base-emitter depletion region.

  17. Effect of high gamma background on neutron sensitivity of fission detectors

    International Nuclear Information System (INIS)

    Balagi, V.; Prasad, K.R.; Kataria, S.K.

    2004-01-01

    Tests were performed on two parallel plate and two cylindrical fission detectors in pulse and dc mode. The effect of gamma background on neutron sensitivity was studied in thermal neutron flux from 30 nv to 60 nv over which gamma field intensity ranging from 230 kR/h to 3.7 MR/h was superposed. In the case of one of the parallel plate detectors the fall in neutron sensitivity was observed to be 3.7% at 1 MR/h and negligible below 1 MR/h. In the case of one of the cylindrical counters the fall in neutron sensitivity was negligible below 500 kR/h and 37% at 1 MR/h. The data was used to derive the design parameters for a wide range fission detector to be procured for PFBR instrumentation for operation at 600 degC and gamma background of 1 MR/h. (author)

  18. Radiation effect on silicon transistors in mixed neutrons-gamma environment

    Science.gov (United States)

    Assaf, J.; Shweikani, R.; Ghazi, N.

    2014-10-01

    The effects of gamma and neutron irradiations on two different types of transistors, Junction Field Effect Transistor (JFET) and Bipolar Junction Transistor (BJT), were investigated. Irradiation was performed using a Syrian research reactor (RR) (Miniature Neutron Source Reactor (MNSR)) and a gamma source (Co-60 cell). For RR irradiation, MCNP code was used to calculate the absorbed dose received by the transistors. The experimental results showed an overall decrease in the gain factors of the transistors after irradiation, and the JFETs were more resistant to the effects of radiation than BJTs. The effect of RR irradiation was also greater than that of gamma source for the same dose, which could be because neutrons could cause more damage than gamma irradiation.

  19. Bismuth- and lithium-loaded plastic scintillators for gamma and neutron detection

    International Nuclear Information System (INIS)

    Cherepy, Nerine J.; Sanner, Robert D.; Beck, Patrick R.; Swanberg, Erik L.; Tillotson, Thomas M.; Payne, Stephen A.; Hurlbut, Charles R.

    2015-01-01

    Transparent plastic scintillators based on polyvinyltoluene (PVT) have been fabricated with high loading of bismuth carboxylates for gamma spectroscopy, and with lithium carboxylates for neutron detection. When activated with a combination of standard fluors, 2,5-diphenyloxazole (PPO) and tetraphenylbutadiene (TPB), gamma light yields with 15 wt% bismuth tripivalate of 5000 Ph/MeV are measured. A PVT plastic formulation including 30 wt% lithium pivalate and 30 wt% PPO offers both pulse shape discrimination, and a neutron capture peak at ~400 keVee. In another configuration, a bismuth-loaded PVT plastic is coated with ZnS( 6 Li) paint, permitting simultaneous gamma and neutron detection via pulse shape discrimination with a figure-of-merit of 3.8, while offering gamma spectroscopy with energy resolution of R(662 keV)=15%

  20. Optimization of electret ionization chambers for dosimetry in mixed neutron-gamma fields

    International Nuclear Information System (INIS)

    Doerschel, B.; Pretzsch, G.

    1984-01-01

    The properties of combination dosemeters consisting of two air-filled electret ionization chambers in mixed neutron-gamma fields have been investigated. The first chamber, polyethylene-walled, is sensitive to neutrons and gamma rays, the second, having walls of teflon, is sensitive to gamma rays only. The properties of the dosemeters are determined by the resulting errors and the measuring range. As both properties depend on the dimensions of the electret ionization chambers they have been taken into account in optimizing the dimensions. The results show that with the use of the dosemeters the effective dose equivalent in mixed neutron-gamma fields can be determined nearly independently of the spectra. The lower detection limit is less than 1 mSv and the maximum uncertainty of dose measurements about 12%. (author)

  1. Nondestructive assay methodologies in nuclear forensics analysis

    International Nuclear Information System (INIS)

    Tomar, B.S.

    2016-01-01

    In the present chapter, the nondestructive assay (NDA) methodologies used for analysis of nuclear materials as a part of nuclear forensic investigation have been described. These NDA methodologies are based on (i) measurement of passive gamma and neutrons emitted by the radioisotopes present in the nuclear materials, (ii) measurement of gamma rays and neutrons emitted after the active interrogation of the nuclear materials with a source of X-rays, gamma rays or neutrons

  2. Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Austin Lewis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Talou, Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stetcu, Ionel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kiedrowski, Brian Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jaffke, Patrick John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-12-06

    CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivity of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.

  3. Neutron induced gamma spectrometry for on-line compositional analysis in coal conversion and fluidized-bed combustion plants

    International Nuclear Information System (INIS)

    Herzenberg, C.L.; O'Fallon, N.M.; Yarlagadda, B.S.; Doering, R.W.; Cohn, C.E.; Porges, K.G.; Duffey, D.

    1977-01-01

    Nuclear techniques involving relatively penetrating radiation may offer the possibility of non-invasive, continuous on-line instrumental monitoring which is representative of the full process stream. Prompt gamma rays following neutron capture are particularly attractive because the penetrating power of the neutrons and the, typically several MeV, capture gammas makes possible interrogation of material within a pipe. We are evaluating neutron capture gamma techniques for this application, both for elemental composition monitoring and for mass-flow measurement purposes, and this paper will present some recent work on composition analysis by neutron induced gamma spectrometry

  4. Application of neutron multiplicity counting to waste assay

    Energy Technology Data Exchange (ETDEWEB)

    Pickrell, M.M.; Ensslin, N. [Los Alamos National Lab., NM (United States); Sharpe, T.J. [North Carolina State Univ., Raleigh, NC (United States)

    1997-11-01

    This paper describes the use of a new figure of merit code that calculates both bias and precision for coincidence and multiplicity counting, and determines the optimum regions for each in waste assay applications. A {open_quotes}tunable multiplicity{close_quotes} approach is developed that uses a combination of coincidence and multiplicity counting to minimize the total assay error. An example is shown where multiplicity analysis is used to solve for mass, alpha, and multiplication and tunable multiplicity is shown to work well. The approach provides a method for selecting coincidence, multiplicity, or tunable multiplicity counting to give the best assay with the lowest total error over a broad spectrum of assay conditions. 9 refs., 6 figs.

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

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

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

  6. {gamma}-Ray background sources in the VESUVIO spectrometer at ISIS spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Pietropaolo, A. [CNISM Milano-Bicocca, Universita degli Studi di Milano-Bicocca, Dipartimento di Fisica ' G. Occhialini' , Piazza della Scienza 3, 20126 Milano (Italy); NAST Center (Nanoscienze-Nanotecnologie-Strumentazione), Universita degli Studi di Roma Tor Vergata, via della Ricerca Scientifica 1, 00133 Roma (Italy)], E-mail: antonino.pietropaolo@mib.infn.it; Perelli Cippo, E. [Universita degli Studi di Milano-Bicocca, Dipartimento di Fisica ' G. Occhialini' , Piazza della Scienza 3, 20126 Milano (Italy); Gorini, G. [CNISM Milano-Bicocca, Universita degli Studi di Milano-Bicocca, Dipartimento di Fisica ' G. Occhialini' , Piazza della Scienza 3, 20126 Milano (Italy); NAST Center (Nanoscienze-Nanotecnologie-Strumentazione), Universita degli Studi di Roma Tor Vergata, via della Ricerca Scientifica 1, 00133 Roma (Italy); Tardocchi, M. [Universita degli Studi di Milano-Bicocca, Dipartimento di Fisica ' G. Occhialini' , Piazza della Scienza 3, 20126 Milano (Italy); Schooneveld, E.M. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire 0QX OX11 (United Kingdom); Andreani, C.; Senesi, R. [Universia degli Studi di Roma Tor Vergata, Dipartimento di Fisica and NAST Center (Nanoscienze-Nanotecnologie-Strumentazione), via della Ricerca Scientifica 1, 00133 Roma (Italy)

    2009-09-01

    An investigation of the gamma background was carried out in the VESUVIO spectrometer at the ISIS spallation neutron source. This study, performed with a yttrium-aluminum-perovskite (YAP) scintillator, follows high resolution pulse height measurements of the gamma background carried out on the same instrument with the use of a high-purity germanium detector. In this experimental work, a mapping of the gamma background was attempted, trying to find the spatial distribution and degree of directionality of the different contributions identified in the previous study. It is found that the gamma background at low times is highly directional and mostly due to the gamma rays generated in the moderator-decoupler system. The other contributions, consistently to the findings of a previous experiment, are identified as a nearly isotropic one due to neutron absorption in the walls of the experimental hall, and a directional one coming from the beam dump.

  7. Tests of the space gamma spectrometer prototype at the JINR experimental facility with different types of neutron generators

    Science.gov (United States)

    Litvak, M. L.; Vostrukhin, A. A.; Golovin, D. V.; Dubasov, P. V.; Zontikov, A. O.; Kozyrev, A. S.; Krylov, A. R.; Krylov, V. A.; Mitrofanov, I. G.; Mokrousov, M. I.; Repkin, A. N.; Timoshenko, G. N.; Udovichenko, K. V.; Shvetsov, V. N.

    2017-07-01

    The results of the tests of the HPGe gamma spectrometer performed with a planetary soil model and different types of pulse neutron generators are presented. All measurements have been performed at the experimental nuclear planetary science facility (Joint Institute for Nuclear Research) for the physical calibration of active gamma and neutron spectrometers. The aim of the study is to model a space experiment on determining the elemental composition of Martian planetary matter by neutron-induced gamma spectroscopy. The advantages and disadvantages of a gas-filled neutron generator in comparison with a vacuum-tube neutron generator are examined.

  8. Prompt gamma activation analysis (PGAA) and short-lived neutron activation analysis (NAA) applied to the characterization of legacy materials

    International Nuclear Information System (INIS)

    English, G.A.; Firestone, R.B.; Perry, D.L.; Reijonen, J.P.; Ka-Ngo Leung; Garabedian, G.F.; Molnar, G.L.; Revay, Zs.

    2008-01-01

    Without quality historical records that provide the composition of legacy materials, the elemental and/or chemical characterization of such materials requires a manual analytical strategy that may expose the analyst to unknown toxicological hazards. In addition, much of the existing legacy inventory also incorporates radioactivity, and, although radiological composition may be determined by various nuclear-analytical methods, most importantly, gamma-spectroscopy, current methods of chemical characterization still require direct sample manipulation, thereby presenting special problems with broad implications for both the analyst and the environment. Alternately, prompt gamma activation analysis (PGAA) provides a 'single-shot' in-situ, non-destructive method that provides a complete assay of all major entrained elemental constituents. Additionally, neutron activation analysis (NAA) using short-lived activation products complements PGAA and is especially useful when NAA activation surpasses the PGAA in elemental sensitivity. (author)

  9. Prompt gamma activation analysis (PGAA) and short-lived neutron activation analysis (NAA) applied to the characterization of legacy materials

    International Nuclear Information System (INIS)

    Firestone, Richard B; English, G.A.; Firestone, R.B.; Perry, D.L.; Reijonen, J.P.; Leung, Ka-Ngo; Garabedian, G.F.; Molnar, G.L.; Revay, Zs.

    2008-01-01

    Without quality historical records that provide the composition of legacy materials, the elemental and/or chemical characterization of such materials requires a manual analytical strategy that may expose the analyst to unknown toxicological hazards. In addition, much of the existing legacy inventory also incorporates radioactivity, and, although radiological composition may be determined by various nuclear-analytical methods, most importantly, gamma-spectroscopy, current methods of chemical characterization still require direct sample manipulation, thereby presenting special problems with broad implications for both the analyst and the environment. Alternately, prompt gamma activation analysis (PGAA) provides a 'single-shot' in-situ, non-destructive method that provides a complete assay of all major entrained elemental constituents.1-3. Additionally, neutron activation analysis (NAA) using short-lived activation products complements PGAA and is especially useful when NAA activation surpasses the PGAA in elemental sensitivity

  10. Investigation of dose distribution in mixed neutron-gamma field of boron neutron capture therapy using N isopropylacrylamide gel

    Energy Technology Data Exchange (ETDEWEB)

    Bavarmegin, Elham; Sadremomtaz, Alireza [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Khalafi, Hossein; Kasesaz, Yaser [Dept. of Physics, University of Guilan, Rasht (Iran, Islamic Republic of); Khajeali, Azim [Medical Education Research Center, Tabriz (Iran, Islamic Republic of)

    2017-02-15

    Gel dosimeters have unique advantages in comparison with other dosimeters. Until now, these gels have been used in different radiotherapy techniques as a reliable dosimetric tool. Because dose distribution measurement is an important factor for appropriate treatment planning in different radiotherapy techniques, in this study, we evaluated the ability of the N-isopropylacrylamide (NIPAM) polymer gel to record the dose distribution resulting from the mixed neutron-gamma field of boron neutron capture therapy (BNCT). In this regard, a head phantom containing NIPAM gel was irradiated using the Tehran Research Reactor BNCT beam line, and then by a magnetic resonance scanner. Eventually, the R2 maps were obtained in different slices of the phantom by analyzing T2-weighted images. The results show that NIPAM gel has a suitable potential for recording three-dimensional dose distribution in mixed neutron-gamma field dosimetry.

  11. Gamma-ray-spectroscopy following high-flux 14-MeV neutron activation

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R.E.

    1981-10-12

    The Rotating Target Neutron Source (RTNS-I), a high-intensity source of 14-MeV neutrons at the Lawrence Livermore National Laboratory (LLNL), has been used for applications in activation analysis, inertial-confinement-fusion diagnostic development, and fission decay-heat studies. The fast-neutron flux from the RTNS-I is at least 50 times the maximum fluxes available from typical neutron generators, making these applications possible. Facilities and procedures necessary for gamma-ray spectroscopy of samples irradiated at the RTNS-I were developed.

  12. Gamma-ray-spectroscopy following high-flux 14-MeV neutron activation

    International Nuclear Information System (INIS)

    Williams, R.E.

    1981-01-01

    The Rotating Target Neutron Source (RTNS-I), a high-intensity source of 14-MeV neutrons at the Lawrence Livermore National Laboratory (LLNL), has been used for applications in activation analysis, inertial-confinement-fusion diagnostic development, and fission decay-heat studies. The fast-neutron flux from the RTNS-I is at least 50 times the maximum fluxes available from typical neutron generators, making these applications possible. Facilities and procedures necessary for gamma-ray spectroscopy of samples irradiated at the RTNS-I were developed

  13. Feasibility study of {sup 235}U and {sup 239}Pu characterization in radioactive waste drums using neutron-induced fission delayed gamma rays

    Energy Technology Data Exchange (ETDEWEB)

    Nicol, T. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Pérot, B., E-mail: bertrand.perot@cea.fr [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Carasco, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Brackx, E. [CEA, DEN, Marcoule, Metallography and Chemical Analysis Laboratory, F-30207 Bagnols-sur-Cèze (France); Mariani, A.; Passard, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Mauerhofer, E. [FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Collot, J. [Laboratoire de Physique Subatomique et de Cosmologie, Université Grenoble Alpes, CNRS/IN2P3 Grenoble (France)

    2016-10-01

    This paper reports a feasibility study of {sup 235}U and {sup 239}Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of {sup 235}U and {sup 239}Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to {sup 137}Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of {sup 235}U or {sup 239}Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.

  14. Production of low energy gamma rays by neutron interactions with fluorine for incident neutron energies between 0.1 and 20 MeV

    International Nuclear Information System (INIS)

    Morgan, G.L.; Dickens, J.K.

    1975-06-01

    Differential cross sections for the production of low-energy gamma rays (less than 240 keV) by neutron interactions in fluorine have been measured for neutron energies between 0.1 and 20 MeV. The Oak Ridge Electron Linear Accelerator was used as the neutron source. Gamma rays were detected at 92 0 using an intrinsic germanium detector. Incident neutron energies were determined by time-of-flight techniques. Tables are presented for the production cross sections of three gamma rays having energies of 96, 110, and 197 keV. (14 figures, 3 tables) (U.S.)

  15. Standard test method for nondestructive assay of radioactive material by tomographic gamma scanning

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method describes the nondestructive assay (NDA) of gamma ray emitting radionuclides inside containers using tomographic gamma scanning (TGS). High resolution gamma ray spectroscopy is used to detect and quantify the radionuclides of interest. The attenuation of an external gamma ray transmission source is used to correct the measurement of the emission gamma rays from radionuclides to arrive at a quantitative determination of the radionuclides present in the item. 1.2 The TGS technique covered by the test method may be used to assay scrap or waste material in cans or drums in the 1 to 500 litre volume range. Other items may be assayed as well. 1.3 The test method will cover two implementations of the TGS procedure: (1) Isotope Specific Calibration that uses standards of known radionuclide masses (or activities) to determine system response in a mass (or activity) versus corrected count rate calibration, that applies to only those specific radionuclides for which it is calibrated, and (2) Respo...

  16. Determination of planetary surfaces elemental composition by gamma and neutron spectroscopy

    International Nuclear Information System (INIS)

    Diez, B.

    2009-06-01

    Measuring the neutron and gamma ray fluxes produced by the interaction of galactic cosmic rays with planetary surfaces allow constraining the chemical composition of the upper tens of centimeters of material. Two different angles are proposed to study neutron and gamma spectroscopy: data processing and data interpretation. The present work is in line with two experiments, the Mars Odyssey Neutron Spectrometer (MONS) and the Selene Gamma Ray Spectrometer. A review of the processing operations applied to the MONS dataset is proposed. The resulting dataset is used to determine the depth of the hydrogen deposits below the Martian surface. In water depleted regions, neutron data allow constraining the concentration in elements likely to interact with neutrons. The confrontation of these results to those issued from the Gamma Ray Spectrometer onboard Mars Odyssey provides interesting insight on the geologic context of the Central Elysium Planitia region. These martian questions are followed by the study of the Selene gamma ray data. Although only preliminary processing has been done to date, qualitative lunar maps of major elements (Fe, Ca, Si, Ti, Mg, K, Th, U) have already been realized. (author)

  17. Synergistic effect of mixed neutron and gamma irradiation in bipolar operational amplifier OP07

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Liu, E-mail: liuyan@nint.ac.cn [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O.Box 69-10, Xi’an 710024 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wei, Chen; Shanchao, Yang; Xiaoming, Jin [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O.Box 69-10, Xi’an 710024 (China); Chaohui, He [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China)

    2016-09-21

    This paper presents the synergistic effects in bipolar operational amplifier OP07. The radiation effects are studied by neutron beam, gamma ray, and mixed neutron/gamma ray environments. The characterateristics of the synergistic effects are studied through comparison of different experiment results. The results show that the bipolar operational amplifier OP07 exhibited significant synergistic effects in the mixed neutron and gamma irradiation. The bipolar transistor is identified as the most radiation sensitive unit of the operational amplifier. In this paper, a series of simulations are performed on bipolar transistors in different radiation environments. In the theoretical simulation, the geometric model and calculations based on the Medici toolkit are built to study the radiation effects in bipolar components. The effect of mixed neutron and gamma irradiation is simulated based on the understanding of the underlying mechanisms of radiation effects in bipolar transistors. The simulated results agree well with the experimental data. The results of the experiments and simulation indicate that the radiation effects in the bipolar devices subjected to mixed neutron and gamma environments is not a simple combination of total ionizing dose (TID) effects and displacement damage. The data suggests that the TID effect could enhance the displacement damage. The synergistic effect should not be neglected in complex radiation environments.

  18. Detection of garlic gamma-irradiated by assay comet

    International Nuclear Information System (INIS)

    Moreno Alvarez, Damaris L.; Miranda, Enrique F. Prieto; Carro, Sandra; Iglesias Enrique, Isora; Matos, Wilberto

    2009-01-01

    The garlic samples were irradiated in a facility with 60 Co sources, at absorbed dose values of 0-0,15 kGy. The detection method utilized for the identification of the irradiated garlic was biological comet assay. The samples were classified post-irradiation several times. The irradiated samples showed high strand breaks of DNA exhibiting comets of several forms, while the not irradiated and lower dose samples showed a behavior like round shape and light comets. Significant differences were found for higher absorbed dose values at 0.06 kGy, this absorbed dose value is corresponding with the applied dose value at this food in order to avoid the germination. (author)

  19. Detection of garlic gamma-irradiated by assay comet

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Alvarez, Damaris L.; Miranda, Enrique F. Prieto; Carro, Sandra; Iglesias Enrique, Isora; Matos, Wilberto [Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear (CEADEN), Ciudad de La Habana (Cuba)], e-mail: damaris@ceaden.edu.cu

    2009-07-01

    The garlic samples were irradiated in a facility with {sup 60}Co sources, at absorbed dose values of 0-0,15 kGy. The detection method utilized for the identification of the irradiated garlic was biological comet assay. The samples were classified post-irradiation several times. The irradiated samples showed high strand breaks of DNA exhibiting comets of several forms, while the not irradiated and lower dose samples showed a behavior like round shape and light comets. Significant differences were found for higher absorbed dose values at 0.06 kGy, this absorbed dose value is corresponding with the applied dose value at this food in order to avoid the germination. (author)

  20. A low-background gamma-ray assay laboratory for activation analysis

    International Nuclear Information System (INIS)

    Lindstrom, R.M.; Langland, J.K.; Lindstrom, D.J.; Slaback, L.A.

    1990-01-01

    The sources of background in a gamma-ray detector were experimentally determined in underground and surface counting rooms, and an optimized shield was constructed at NIST. The optimum thickness of lead was 10-15 cm, with a greater thickness giving an increased background due to the buildup of tertiary cosmic-ray particles. Neither cadmium, tin, copper nor plastic (hydrocarbon or fluorocarbon) was desirable as a shield liner, since all these increased the background continuum or introduced characteristic peaks into the background spectrum. Two broad peaks in the background result from inelastic scattering of cosmic-ray neutrons (0.02 cm -2 s -1 ) in germanium. These neutrons also excite the lower nuclear levels of lead and structural iron to produce additional gamma-ray peaks in the spectrum. The influence of the 20 MW NIST reactor, located 60 m from the detector, was undetectable. Comparisons among detectors and locations clearly separate cosmic from environmental components of the background. (orig.)

  1. Reduction of bias in neutron multiplicity assay using a weighted point model

    Energy Technology Data Exchange (ETDEWEB)

    Geist, W. H. (William H.); Krick, M. S. (Merlyn S.); Mayo, D. R. (Douglas R.)

    2004-01-01

    Accurate assay of most common plutonium samples was the development goal for the nondestructive assay technique of neutron multiplicity counting. Over the past 20 years the technique has been proven for relatively pure oxides and small metal items. Unfortunately, the technique results in large biases when assaying large metal items. Limiting assumptions, such as unifoh multiplication, in the point model used to derive the multiplicity equations causes these biases for large dense items. A weighted point model has been developed to overcome some of the limitations in the standard point model. Weighting factors are detemiined from Monte Carlo calculations using the MCNPX code. Monte Carlo calculations give the dependence of the weighting factors on sample mass and geometry, and simulated assays using Monte Carlo give the theoretical accuracy of the weighted-point-model assay. Measured multiplicity data evaluated with both the standard and weighted point models are compared to reference values to give the experimental accuracy of the assay. Initial results show significant promise for the weighted point model in reducing or eliminating biases in the neutron multiplicity assay of metal items. The negative biases observed in the assay of plutonium metal samples are caused by variations in the neutron multiplication for neutrons originating in various locations in the sample. The bias depends on the mass and shape of the sample and depends on the amount and energy distribution of the ({alpha},n) neutrons in the sample. When the standard point model is used, this variable-multiplication bias overestimates the multiplication and alpha values of the sample, and underestimates the plutonium mass. The weighted point model potentially can provide assay accuracy of {approx}2% (1 {sigma}) for cylindrical plutonium metal samples < 4 kg with {alpha} < 1 without knowing the exact shape of the samples, provided that the ({alpha},n) source is uniformly distributed throughout the

  2. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  3. Recognition of internal structure of unknown objects with simultaneous neutron and gamma radiography

    International Nuclear Information System (INIS)

    Moghadam, K.K.; Nasseri, M.M.

    2004-01-01

    Generally speaking in customary industrial and medical radiography, there is no tendency to reveal the nature of the samples. Ordinarily, the main objective of taking a radiograph is to show the position and dimension of unknown parts, inside the test object and to determine cracks, defects, etc. Whereas in radiography many important factors such as material cross-sections and build-up factors are also involved. In this paper, by using both neutron and gamma radiography techniques, some mathematical relations were successfully generated, in order to calculate the neutron and gamma total macroscopic cross-sections of some unknown elements in the presence of the other elements. For this work, some test pieces were defined and made of lead, silver, copper, Nickel, tin, graphite and polyethylene. The neutron radiography facility at Tehran Research Reactor (TRR) was used as mixed neutron and gamma radiography source (Proceedings of the Second World Conference on Neutron Radiography, Paris, France, pp. 25-32). On testing of a correction of the above-mentioned generated relations, a new technique of simultaneous neutron and gamma radiography was also investigated

  4. A Unique Outside Neutron and Gamma Ray Instrumentation Development Test Facility at NASA's Goddard Space Flight Center

    Science.gov (United States)

    Bodnarik, J.; Evans, L.; Floyd, S.; Lim, L.; McClanahan, T.; Namkung, M.; Parsons, A.; Schweitzer, J.; Starr, R.; Trombka, J.

    2010-01-01

    An outside neutron and gamma ray instrumentation test facility has been constructed at NASA's Goddard Space Flight Center (GSFC) to evaluate conceptual designs of gamma ray and neutron systems that we intend to propose for future planetary lander and rover missions. We will describe this test facility and its current capabilities for operation of planetary in situ instrumentation, utilizing a l4 MeV pulsed neutron generator as the gamma ray excitation source with gamma ray and neutron detectors, in an open field with the ability to remotely monitor and operate experiments from a safe distance at an on-site building. The advantage of a permanent test facility with the ability to operate a neutron generator outside and the flexibility to modify testing configurations is essential for efficient testing of this type of technology. Until now, there have been no outdoor test facilities for realistically testing neutron and gamma ray instruments planned for solar system exploration

  5. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. M.; Sum, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified.

  6. A dual neutron/gamma source for the Fissmat Inspection for Nuclear Detection (FIND) system.

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); King, Michael; Rossi, Paolo (Sandia National Laboratories, Albuquerque, NM); McDaniel, Floyd Del (Sandia National Laboratories, Albuquerque, NM); Morse, Daniel Henry; Antolak, Arlyn J.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM); Raber, Thomas N.

    2008-12-01

    Shielded special nuclear material (SNM) is very difficult to detect and new technologies are needed to clear alarms and verify the presence of SNM. High-energy photons and neutrons can be used to actively interrogate for heavily shielded SNM, such as highly enriched uranium (HEU), since neutrons can penetrate gamma-ray shielding and gamma-rays can penetrate neutron shielding. Both source particles then induce unique detectable signals from fission. In this LDRD, we explored a new type of interrogation source that uses low-energy proton- or deuteron-induced nuclear reactions to generate high fluxes of mono-energetic gammas or neutrons. Accelerator-based experiments, computational studies, and prototype source tests were performed to obtain a better understanding of (1) the flux requirements, (2) fission-induced signals, background, and interferences, and (3) operational performance of the source. The results of this research led to the development and testing of an axial-type gamma tube source and the design/construction of a high power coaxial-type gamma generator based on the {sup 11}B(p,{gamma}){sup 12}C nuclear reaction.

  7. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    International Nuclear Information System (INIS)

    Lee, K. M.; Sum, G. M.

    2016-01-01

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified

  8. Assay of fissionable isotopes in aqueous solution by pulsed neutron interrogation

    International Nuclear Information System (INIS)

    Campbell, P.; Gardy, E.M.; Boase, D.G.

    1978-04-01

    Non-destructive assay of uranium-235 and thorium-232 in aqueous nitric acid solutions has been accomplished by irradiation with pulses of neutrons from a 14-MeV Cockcroft-Walton neutron generator, and counting of the delayed neutrons emitted from the fissions induced. Design of the delayed neutron detector assemblies is described, together with the neutron pulse timing and counting systems. The effects of irradiation time, counting time, neutron moderation, detector design and sample geometry on the delayed neutron response from uranium-235 and 238 and thorium-232 are discussed. By using polyethylene to moderate the interrogating neutrons, solutions can be analyzed for both uranium-235 and thorium. Comparative analyses with chemical and γ-spectrometric methods show good agreement. The neutron method is rapid and is shown to be unaffected by the presence in solution of impurities such as iron, nickel, chromium, and aluminum. With the experimental equipment described, detection limits of 0.6 mg of 235 U and 9 mg of 232 Th in a sample volume of 25 mL have been achieved. Analyses of highly radioactive samples may be done easily since the measurements are not affected by the presence of large amounts of βγ radiation. Samples can be enclosed in small lead-shielded flasks during analysis to protect the analyst. The potential of the technique to on-line analysis applications is explored briefly. (author)

  9. Multi-isotopic gamma-ray assay system for alpha-contaminated waste

    International Nuclear Information System (INIS)

    Close, D.A.; Pratt, J.C.; Caldwell, J.T.; Kunz, W.E.; Schultz, F.J.; Haff, K.W.

    1983-01-01

    The capability of an existing segmented gamma-ray system is being expanded for the analysis of alpha-contaminated waste drums. A cursory assay of 114 transuranic waste drums of 208-l capacity has been made. Analysis of these data indicates a detection limit better than 100 nCi/g of waste for 237 Np/ 233 Pa, 239 Pu, 241 Am, 243 Am/ 239 Np, 60 Co, 125 Sb, 134 137 Cs, and 154 Eu. A pending Code of Federal Regulation (10CFR61) stipulates that the nuclear industry quantify not only its transuranic waste, but also certain beta- and gamma-ray-emitting fission products. An assay system based on gamma-ray spectroscopy is the only system that can meet this requirement for the fission products

  10. Non-destructive assay of fissile materials by detection and multiplicity analysis of spontaneous neutrons

    International Nuclear Information System (INIS)

    Prosdocimi, A.

    1979-01-01

    A method for determining the absolute reaction rate of nuclear events giving rise to neutron emission, according to their neutron multiplicity, is proposed. A typical application is the measurement of the (α, n) and spontaneous fission rates in a fissile material sample, particularly of Pu oxide composition. An analysis of random and correlated neutron pulses is carried out on the basis of sequential order without requiring any time interval analysis, then the primary nuclear events are sorted versus their neutron multiplicity. Suitable theoretical relationships enable to derive the absolute (α, n) and SF reaction rates when the physical parameters of the neutron detector and the multiplicity spectrumm of pulses are known. A typical device is described and the results of experiments leading to Pu-239 and Pu-240 assay are given

  11. Capability and limitation study of the DDT passive-active neutron waste assay instrument

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Coop, K.L.; Estep, R.J.

    1992-05-01

    The differential-dieaway-technique passive-active neutron assay system is widely used by transuranic waste generators to certify their drummed waste for eventual shipment to the Waste Isolation Pilot Plant (WIPP). Stricter criteria being established for waste emplacement at the WIPP site has led to a renewed interest in improvements to and a better understanding of current nondestructive assay (NDA) techniques. Our study includes the effects of source position, extreme matrices, high neutron backgrounds, and source self-shielding to explore the system's capabilities and limitations and to establish a basis for comparison with other NDA systems. 11 refs

  12. Device for characterization of fissile materials comprising at least a neutron detector embedded inside a scintillator for gamma radiation detection

    International Nuclear Information System (INIS)

    Bernard, P.; Dherbey, J.R.; Bosser, R.; Berne, R.

    1989-01-01

    Fissile materials, for instance in radioactive wastes, are characterized by measurement of prompt and delayed neutrons and gamma radiation from induced fission by a neutron source. Gamma radiation is detected with a scintillation detector associated to a photomultiplier, the scintillation material is at the same time a moderator for thermalization of fast neutrons emitted by the neutron source and also of neutrons from spontaneous fission, (α, n) reactions and neutrons from induced fission in the fissile material. Preferentially the moderator is made of Altustipe (Plexiglas with anthracene as additive) [fr

  13. Gamma-ray spectroscopy of neutron-rich products of heavy-ion collisions

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, M.P.; Janssens, R.V.F.; Ahmad, I. [and others

    1995-08-01

    Thick-target {gamma}{gamma} coincidence techniques are being used to explore the spectroscopy of otherwise hard-to-reach neutron-rich products of deep-inelastic heavy ion reactions. Extensive {gamma}{gamma} coincidence measurements were performed at ATLAS using pulsed beams of {sup 80}Se, {sup 136}Xe, and {sup 238}U on lead-backed {sup 122,124}Sn targets with energies 10-15% above the Coulomb barrier. Gamma-ray coincidence intensities were used to map out yield distributions with A and Z for even-even product nuclei around the target and around the projectile. The main features of the yield patterns are understandable in terms of N/Z equilibration. We had the most success in studying the decays of yrast isomers. Thus far, more than thirty new {mu}s isomers in the Z = 50 region were found and characterized. Making isotopic assignments for previously unknown {gamma}-ray cascades proves to be one of the biggest problems. Our assignments were based (a) on rare overlaps with radioactivity data, (b) on the relative yields with different beams, and (c) on observed cross-coincidences between {gamma} rays from light and heavy reaction partners. However, the primary products of deep inelastic collisions often are sufficiently excited for subsequent neutron evaporation, so {gamma}{gamma} cross-coincidence results require careful interpretation.

  14. Calculation of neutron and gamma-ray energy spectra in liquid air and liquid nitrogen due to 14-MeV neutron and californium-252 sources

    International Nuclear Information System (INIS)

    Straker, E.A.; Gritzner, M.L.; Harris, L. Jr.

    1978-01-01

    Calculations of neutron and gamma-ray fluences from 14-MeV neutron and 252 Cf sources in liquid air and liquid nitrogen have been performed. These calculations were made specifically for comparison with experimental data measured at Stohl, Federal Republic of Germany. The discrete-ordinates method was utilized with neutron and gamma-ray cross sections from ENDF/B-IV. One-dimensional calculational models were developed for the sources and tank. Limited comparisons are made with experimental data

  15. Neutron and gamma ray attenuation of asphalt; Comparison with paraffin and water

    International Nuclear Information System (INIS)

    Abdul-Majid, S.; Kutbi, I.I.

    1996-01-01

    Asphalt is a low cost, readily available, easy-to-cast material which is rich in hydrogen and carbon, elements most effective for fast-neutron shielding. Unlike paraffin, the material can easily be mixed with boron containing compounds, an, element of high absorption cross-section for slow neutrons. The 241 Am-Be neutron and gamma attenuation characteristic of asphalt were studied. The source is having wide applications in industry and geophysics field work. Comparisons were made with paraffin and water. The source activity was 1.11 x 1,011 Bq (3 Ci) with a neutron emission rate of 6.6 x 106 n s -1 and a tolerance of +10%. The neutron dose-equivalent rate at 1 m was 66 mSv h -1 , while the associated gamma ray exposure was ∼1.9 mC kg -1 h -1 of the bare source. A neutron remmeter was used for the neutron dose-equivalent rate measurements, which produces an energy response that approximates human body dose equivalent over a wide range of neutron energy. An air filled ionization chamber was used for the exposure rate measurements. The slow neutrons were measured by a BF 3 gas filled detector. The shielding materials were confined in an aluminum cylinder of 1 mm wall thickness where the source was kept in the middle. The neutron dose rate, the gamma ray exposure rate, and the slow neutron count rate were measured at different shield radii and at different distances from its outer surface. The neutron doses of asphalt at the surface of cylindrical shields of 8, 12, 16, 20, and 24 cm radii in mSv h -1 were 0.85, 0.4, 0.25, 0.13, and 0.06, respectively, while the gamma ray exposure mC kg -1 h -1 were 7, 4.4 2.5, 1.3, and 0.88, respectively. The neutron dose rate attenuation of asphalt was very close to that of water, but slightly lower than that of paraffin, while the gamma ray attenuation was close to that of water but higher than that of paraffin

  16. Bulk media assay using backscattered Pu-Be neutrons

    CERN Document Server

    Csikai, J

    1999-01-01

    Spectral yields of elastically backscattered Pu-Be neutrons measured for graphite, water, polyethylene, liquid nitrogen, paraffin oil, SiO sub 2 , Al, Fe, and Pb slabs show a definite correlation with the energy dependence of the elastic scattering cross sections, sigma sub E sub L (E sub n). The C, N and O can be identified by the different structures in their sigma sub E sub L (E sub n) functions. The integrated spectral yields versus thickness exhibit saturation for each sample. The interrogated volume is limited by the presence of hydrogen in the sample. (author)

  17. Neutron Resonance Transmission Analysis (NRTA): A Nondestructive Assay Technique for the Next Generation Safeguards Initiative’s Plutonium Assay Challenge

    Energy Technology Data Exchange (ETDEWEB)

    J. W. Sterbentz; D. L. Chichester

    2010-12-01

    This is an end-of-year report for a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The first-year goals for this project were modest and included: 1) developing a zero-order MCNP model for the NRTA technique, simulating data results presented in the literature, 2) completing a preliminary set of studies investigating important design and performance characteristics for the NRTA measurement technique, and 3) documentation of this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes a nine month period of work.

  18. Neutron Resonance Transmission Analysis (NRTA): Initial Studies of a Method for Assaying Plutonium in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David L. Chichester; James W. Sterbentz

    2011-05-01

    Neutron Resonance Transmission Analysis (NRTA) is an analytical technique that uses neutrons to assay the isotopic content of bulk materials. The technique uses a pulsed accelerator to produce an intense, short pulse of neutrons in a time-of-flight configuration. These neutrons, traveling at different speeds according to their energy, can be used to interrogate a spent fuel (SF) assembly to determine its plutonium content. Neutron transmission through the assembly is monitored as a function of neutron energy (time after the pulse), similar to the way neutron cross-section data is often collected. The transmitted neutron intensity is recorded as a function of time, with faster (higher-energy) neutrons arriving first and slower (lower-energy) neutrons arriving later. The low-energy elastic scattering and absorption resonances of plutonium and other isotopes modulate the transmitted neutron spectrum. Plutonium content in SF can be determined by analyzing this attenuation. Work is currently underway at Idaho National Laboratory, as a part of United States Department of Energy's Next Generation Safeguards Initiative (NGSI), to investigate the NRTA technique and to assess its feasibility for quantifying the plutonium content in SF and for determining the diversion of SF pins from assemblies. Preliminary results indicate that NRTA has great potential for being able to assay intact SF assemblies. Operating in the 1-40 eV range, it can identify four plutonium isotopes (239, 240, 241, & 242Pu), three uranium isotopes (235, 236, & 238U), and six resonant fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm). It can determine the areal density or mass of these isotopes in single- or multiple-pin integral transmission scans. Further, multiple observables exist to allow the detection of material diversion (pin defects) including fast-neutron and x-ray radiography, gross-transmission neutron counting, plutonium resonance absorption analysis, and fission

  19. Ornithogalum virens as a plant assay for beta and gamma radiation effects

    International Nuclear Information System (INIS)

    Herron, V.J.

    1979-01-01

    The purpose of this study was to determine if the monocotyledonous angiosperm, Ornithogalum virens (Quintanilha and Cabral, 1947), could be used in such a biological assay system. After exposing O. virens plants to acute ( 60 Co) and chronic ( 137 Cs) gamma radiation and internal beta radiation ( 32 P), lethality (LD 50 , LD 100 ), growth inhibition, and chromosome aberrations were investigated. The LD 50 and LD 100 for acute gamma radiation were estimated to be between 0.91 to 1.8 krad and less than 3.6 krad, respectively. Though growth inhibition and abnormal growth were observed in the acute and chronic gamma radiation studies, the changes in the growth of the plants were so variable that these parameters were found to be unreliable measures of radiation effects. Chromosome aberrations were a more reliable measure of radiation damage because linear relationships between total aberrations and dose were found for both gamma and beta radiation

  20. Prompt-gamma neutron activation analysis system design: Effects of D-T versus D-D neutron generator source selection

    Science.gov (United States)

    Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV and D-T with...

  1. Total Measurement Uncertainty for the Plutonium Finishing Plant (PFP) Segmented Gamma Scan Assay System

    CERN Document Server

    Fazzari, D M

    2001-01-01

    This report presents the results of an evaluation of the Total Measurement Uncertainty (TMU) for the Canberra manufactured Segmented Gamma Scanner Assay System (SGSAS) as employed at the Hanford Plutonium Finishing Plant (PFP). In this document, TMU embodies the combined uncertainties due to all of the individual random and systematic sources of measurement uncertainty. It includes uncertainties arising from corrections and factors applied to the analysis of transuranic waste to compensate for inhomogeneities and interferences from the waste matrix and radioactive components. These include uncertainty components for any assumptions contained in the calibration of the system or computation of the data. Uncertainties are propagated at 1 sigma. The final total measurement uncertainty value is reported at the 95% confidence level. The SGSAS is a gamma assay system that is used to assay plutonium and uranium waste. The SGSAS system can be used in a stand-alone mode to perform the NDA characterization of a containe...

  2. The effects of gamma irradiation on neutron displacement sensitivity of lateral PNP bipolar transistors

    International Nuclear Information System (INIS)

    Wang, Chenhui; Chen, Wei; Liu, Yan; Jin, Xiaoming; Yang, Shanchao; Qi, Chao

    2016-01-01

    The effects of gamma irradiation on neutron displacement sensitivity of four types of lateral PNP bipolar transistors (LPNPs) with different neutral base widths, emitter widths and the doping concentrations of the epitaxial base region are studied. The physical mechanisms of the effects are explored by defect analysis using deep level transient spectroscopy (DLTS) techniques and numerical simulations of recombination process in the base region of the lateral PNP bipolar transistors, and are verified by the experiments on gate-controlled lateral PNP bipolar transistors (GCLPNPs) manufactured in the identical commercial bipolar process with different gate bias voltage. The results indicate that gamma irradiation increases neutron displacement damage sensitivity of lateral PNP bipolar transistors and the mechanism of this phenomenon is that positive charge induced by gamma irradiation enhances the recombination process in the defects induced by neutrons in the base region, leading to larger recombination component of base current and greater gain degradation.

  3. Attenuation of neutrons and gamma-rays in homogeneous and multilayered shields

    International Nuclear Information System (INIS)

    Abdo, A.E.; Megahid, R.M.

    1997-01-01

    Measurements were carried-out to compare the attenuation properties of homogeneous shields and shields of two layers and three layers for fast neutrons and total gamma-rays. These were performed by measuring the fast neutron and total gamma-ray spectra behind homogeneous shields of magnetite-limonite, ilmenite-ilmenite and magnetite-magnetite concretes. The two layers assembly consists of iron and one of the above mentioned concretes, while the three layers shield consists of water, iron and one of the previously mentioned concretes. All measurements were carried-out using a neutron-gamma spectrometer with stilbene scintillator coupled to a fast photo multi player tube. Separation between pulses of recoil protons and recoil electrons was achieved by a pulse shape discrimination technique. 3 tabs., 10 figs., 13 refs

  4. The effects of gamma irradiation on neutron displacement sensitivity of lateral PNP bipolar transistors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chenhui, E-mail: wangchenhui@nint.ac.cn; Chen, Wei; Liu, Yan; Jin, Xiaoming; Yang, Shanchao; Qi, Chao

    2016-09-21

    The effects of gamma irradiation on neutron displacement sensitivity of four types of lateral PNP bipolar transistors (LPNPs) with different neutral base widths, emitter widths and the doping concentrations of the epitaxial base region are studied. The physical mechanisms of the effects are explored by defect analysis using deep level transient spectroscopy (DLTS) techniques and numerical simulations of recombination process in the base region of the lateral PNP bipolar transistors, and are verified by the experiments on gate-controlled lateral PNP bipolar transistors (GCLPNPs) manufactured in the identical commercial bipolar process with different gate bias voltage. The results indicate that gamma irradiation increases neutron displacement damage sensitivity of lateral PNP bipolar transistors and the mechanism of this phenomenon is that positive charge induced by gamma irradiation enhances the recombination process in the defects induced by neutrons in the base region, leading to larger recombination component of base current and greater gain degradation.

  5. Neutron, gamma ray, and temperature effects on the electrical characteristics of thyristors

    Science.gov (United States)

    Frasca, A. J.; Schwarze, G. E.

    1992-01-01

    Experimental data showing the effects of neutrons, gamma rays, and temperature on the electrical and switching characteristics of phase-control and inverter-type SCR's are presented. The special test fixture built for mounting, heating, and instrumenting the test devices is described. Four SCR's were neutron irradiated at 300 K and four at 365 K for fluences up to 3.2 x 10 exp 13 pn/sq. cm, and eight were gamma irradiated at 300 K only for gamma doses up to 5.1 Mrads. The electrical measurements were made during irradiation and the switching measurements were made only before and after irradiation. Radiation induced crystal defects, resulting primarily from fast neutrons, caused the reduction of minority carrier lifetime through the generation of R-G centers. The reduction in lifetime caused increases in the on-state voltage drop and in the reverse and forward leakage currents, and decreases in the turn-off time.

  6. Neutron, gamma ray, and temperature effects on the electrical characteristics of thyristors

    International Nuclear Information System (INIS)

    Schwarze, G.E.; Frasca, A.J.

    1992-01-01

    In this paper, experimental data showing the effects of neutrons, gamma rays, and temperature on the electrical and switching characteristics of phase-control and inverter-type SCRs are presented. The special test fixture built for mounting, heating, and instrumenting the test devices is described. Four SCRs were neutron irradiated at 300 K and four at 365 K for fluences up to 3.2 x 10 13 n/cm 2 , and eight were gamma irradiated at 300 K only for gamma doses up to 5.1 Mrads. The electrical measurements were made during irradiation and the switching measurements were made only before and after irradiation. Radiation induced crystal defects, resulting primarily from fast neutrons, caused the reduction of minority carrier lifetime through the generation of R-G centers. The reduction in lifetime caused increases in the on-state voltage drop and in the reverse and forward leakage currents, and decreases in the turn-off time

  7. High dose effect of gamma and neutrons on the N-JFET electronic components

    International Nuclear Information System (INIS)

    Assaf, Jamal-Eddin

    2006-11-01

    Two types of N-JFET components have been irradiated by high doses of thermal neutrons and gamma rays up to 2000x10 12 n/cm 2 and 1000 kGy, respectively. The static tests show a decrease of the g m and I d s parameters. The behaviour of electronic noise on the output was the principal dynamic test after irradiation. The result of this test gives an increase of the noise with radiation dose increasing. The noise was described as the Equivalent Noise of Charge (ENC) at the output of the measurements set-up. The quantities and the qualities of the noise depend on the N-JEET type and the type of radiation (neutrons or gamma). Other tests were carried out like the relaxation or recovery phenomena after radiation, and the superposed effects of gamma and neutrons.(author)

  8. Performance of neutron and gamma personnel dosimetry in mixed radiation fields

    International Nuclear Information System (INIS)

    Swaja, R.E.; Sims, C.S.

    1981-01-01

    From 1974 to 1980, six personnel dosimetry intercomparison studies (PDIS) were conducted at the Oak Ridge National Laboratory (ORNL) to evaluate the performance of personnel dosimeters in a variety of neutron and gamma fields produced by operating the Health Physics Research Reactor (HPRR) in the steady state mode with and without spectral modifying shields. A total of 58 different organizations participated in these studies which produced approximately 2000 measurements of neutron and gamma dose equivalents on anthropomorphic phantoms for five different reactor spectra. Based on these data, the relative performance of three basic types of neutron dosimeters [nuclear emulsion film, thermoluminescent (TLD), and track-etch] and two basic types of gamma dosimeters (film and TLD) in mixed radiation fields was assessed

  9. Development of automatic gamma and neutron monitoring system for PFBR fuel subassemblies at IFSB

    International Nuclear Information System (INIS)

    Krishnakumar, D.N.; Dhanasekaran, A.; Ajoy, K.C.; Jose, M.T.; Baskaran, R.; Sureshkumar, K.V.

    2018-01-01

    Health physics surveillance during PFBR fuel pin assembling operation at Interim Fuel Storage Building (IFSB) mandates scanning of the fuel assembly using Telector and Rem counter to find out the maximum gamma and neutron dose rates respectively. Throughout the process health physicist involved in the operation must hold the survey meter at a constant distance from the subassembly and simultaneously should make a note of dose rate values displayed. This practice might lead to the occupational exposures and also might induce human errors during measurements. To make this process more simple, faultless and effortless, an automatic Gamma Neutron Monitoring System (AGNMS) is designed and developed at RSD to measure, store and visualize instantaneous gamma and neutron dose rates of PFBR fuel subassembly. Development of the system, calibration and deployment of the system at IFSB and preliminary results obtained using the system is depicted in this paper

  10. Neutron and gamma-ray dose-rates from the Little Boy replica

    International Nuclear Information System (INIS)

    Plassmann, E.A.; Pederson, R.A.

    1984-01-01

    We report dose-rate information obtained at many locations in the near vicinity of, and at distances out to 0.64 km from, the Little Boy replica while it was operated as a critical assembly. The measurements were made with modified conventional dosimetry instruments that used an Anderson-Braun detector for neutrons and a Geiger-Mueller tube for gamma rays with suitable electronic modules to count particle-induced pulses. Thermoluminescent dosimetry methods provide corroborative data. Our analysis gives estimates of both neutron and gamma-ray relaxation lengths in air for comparison with earlier calculations. We also show the neutron-to-gamma-ray dose ratio as a function of distance from the replica. Current experiments and further data analysis will refine these results. 7 references, 8 figures

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. Effects of neutron-gamma or gamma irradiations on plasma clotting factors. Effect of a treatment by substituted factors

    International Nuclear Information System (INIS)

    Mestries, J.C.; Martin, S.; Janodet, D.; Herodin, F.; Gourmelon, P.; Fatome, M.

    1991-01-01

    Neutron-gamma irradiation of the baboon at lethal dose altered the plasma clotting factors and induced a fibrinoformation alteration which occurred shortly before death. These disturbances, which were not found after gamma irradiation, could explain the importance of the haemorrhagic syndrome. Treatment by P.P.S.B. (factors II, VII, X and IX) counteracted the alterations of the plasma clotting factors, but had no influence on the lethality nor on the fibrinoformation alteration which seems to be an important cause of death [fr

  13. Uses of neutron capture gamma-rays in environmental pollution applications

    International Nuclear Information System (INIS)

    AbdAl-Samad, M.A.

    1998-01-01

    As a sensitive and accurate technique, the prompt gamma-rays neutron activation is used with success for elemental analysis. The advantages of this method over the other techniques are rapidity, usage of relatively large sample size and high reliability, beside the detection of the elements which have no gamma activity during the delayed neutron activation analysis or very short lived isotopes. Actually different techniques could be used for estimating the trace, minor and major elements of these environmental samples which are considered as complex samples. In the mean time the neutron activation analysis techniques have been improved and have become an excellent tool for elemental analysis of complex samples (Duffey et al., 1970; Senftle et al., 1971; Henkelmm and Born, 1973 ; Hassan et al., .; 1981, 1982, 1983; Clyton et al., 1983; Zaghloul et al., 1993) and the advantages of the prompt γ- ray neutron activation analysis over the other techniques put this technique in the fore front

  14. Cold neutron prompt gamma activation analysis at NIST; A progress report

    Energy Technology Data Exchange (ETDEWEB)

    Paul, R L; Lindstrom, R M [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Div. of Inorganic Analytical Research; Vincent, D H [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering

    1994-05-01

    An instrument for prompt gamma-ray activation analysis is now in operation at the NIST Cold Neutron Research Facility (CNRF). The cold neutron beam is relatively free of contamination by fast neutrons and reactor gamma rays, and the neutron fluence rate is 1.5 x 10 [sup 8] cm [sup -2] x s [sup -1] (thermal equivalent). As a result of a compact target-detector geometry the sensitivity is better by a factor of as much as seven than that obtained with an existing thermal instrument, and hydrogen background is a factor of 50 lower. This instrument was applied to multielement analysis of the Allende meteorite and other materials. (author) 14 refs.; 2 figs.; 1 tab.

  15. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

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

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  16. Gamma compensated pulsed ionization chamber wide range neutron/reactor power measurement system

    International Nuclear Information System (INIS)

    Ellis, W.H.

    1975-01-01

    An improved method and system of pulsed mode operation of ionization chambers is described in which a single sensor system with gamma compensation is provided by sampling, squaring, automatic gate selector, and differential amplifier circuit means, employed in relation to chambers sensitized to neutron plus gamma and gamma only to subtract out the gamma component, wherein squaring functions circuits, a supplemental high performance pulse rate system, and operational and display mode selection and sampling gate circuits are utilized to provide automatic wide range linear measurement capability for neutron flux and reactor power. Neon is employed as an additive in the ionization chambers to provide independence of ionized gas kinetics temperature effects, and the pulsed mode of operation provide independence of high temperature insulator leakage effects. (auth)

  17. A silicon photomultiplier readout for time of flight neutron spectroscopy with {gamma}-ray detectors

    Energy Technology Data Exchange (ETDEWEB)

    Pietropaolo, A.; Gorini, G. [Dipartimento di Fisica ' ' G. Occhialini' ' and CNISM, Universita Degli Studi di Milano-Bicocca, Piazza della Scienza 3, 20126 Milano (Italy); Festa, G.; Andreani, C.; De Pascale, M. P.; Reali, E. [Dipartimento di Fisica and Centro NAST, Universita degli Studi di Roma Tor Vergata, Via della Ricerca Scientifica 1, 00133, Roma (Italy); Grazzi, F. [Istituto dei Sistemi Complessi-Consiglio Nazionale delle Ricerche, Via Madonna del Piano n.10, I-50019 Sesto Fiorentino, Firenze (Italy); Schooneveld, E. M. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire, OX11 0QX (United Kingdom)

    2009-09-15

    The silicon photomultiplier (SiPM) is a recently developed photosensor used in particle physics, e.g., for detection of minimum ionizing particles and/or Cherenkov radiation. Its performance is comparable to that of photomultiplier tubes, but with advantages in terms of reduced volume and magnetic field insensitivity. In the present study, the performance of a gamma ray detector made of an yttrium aluminum perovskite scintillation crystal and a SiPM-based readout is assessed for use in time of flight neutron spectroscopy. Measurements performed at the ISIS pulsed neutron source demonstrate the feasibility of {gamma}-detection based on the new device.

  18. Synergistic interaction between the neutron and gamma radiation on LACA mice hemopoietic stem cells

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H

    1982-02-01

    Based on the radiation action dual theory of DNA single and double strand breaks, a hypothetical RBE mathematical model for the effect of the mixed radiation of neutron and gamma rays on LACA mice hemopoietic stem cells was formulated. In comparison of the RBE values of different ratio of neutron and gamma-ray mixed radiation with their theoretical additive RBE values, the preliminary impression is that the mixed radiation is more effective than that of the theoretical additive effect. It seems that the existence of synergist in the mixed radiation might be valid.

  19. Neutron-gamma discrimination based on bipolar trapezoidal pulse shaping using FPGAs in NE213

    Energy Technology Data Exchange (ETDEWEB)

    Esmaeili-sani, Vahid, E-mail: vaheed_esmaeely80@yahoo.com [Department of Nuclear Engineering and Physics, Amirkabir University of Technology, P.O. Box 4155-4494, Tehran (Iran, Islamic Republic of); Moussavi-zarandi, Ali; Akbar-ashrafi, Nafiseh; Boghrati, Behzad; Afarideh, Hossein [Department of Nuclear Engineering and Physics, Amirkabir University of Technology, P.O. Box 4155-4494, Tehran (Iran, Islamic Republic of)

    2012-12-01

    A technique employing neutron-gamma pulse shape discrimination (PSD) system that overcomes pile up limitations of previous methods to distinguish neutrons from gammas in scintillation detectors is described. The output signals of detectors were digitized and processed with a data acquisition system based on bipolar trapezoidal pulse shaping using Field programmable gate arrays (FPGA). FPGAs are capable of doing complex discrete signal processing algorithms with clock rates above 100 MHz. Their low cost, ease of use and selected dedicated hardware make them an ideal option for spectrometer systems.

  20. Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma-ray mixed fields

    International Nuclear Information System (INIS)

    Ogawa, Yoshihiro

    1985-01-01

    In order to consider the problems concerned with personnel dosimetry using film badges and TLDs, an intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in five different neutron and gamma-ray mixed fields at KUR and UTR-KINKI from the practical point of view. For the estimation of thermal neutron dose equivalent, it may be concluded that each personnel dosimeter has good performances in the precision, that is, the standard deviations in the measured values by individual dosimeter were within 24 %, and the dose equivalent responses to thermal neutron were almost independent on cadmium ratio and gamma-ray contamination. However, the relative thermal neutron dose equivalent of individual dosimeter normalized to the ICRP recommended value varied considerably and a difference of about 4 times was observed among the dosimeters. From the results obtained, it is suggested that the standardization of calibration factors and procedures is required from the practical point of radiation protection and safety. (author)

  1. Evaluation of gamma radiation induced genetic damage in the fish Cyprinus carpio using comet assay

    International Nuclear Information System (INIS)

    Praveen Kumar, M.K.; Shyama, S.K.; Bhagat, S.S.; Chaubey, R.C.

    2013-01-01

    Radionuclides released from various sources including the industries, as well as, accidental release during a nuclear disaster can contaminate inland water bodies. Suitable bio-monitoring methods/biomarkers are the need of the day to assess the impact of high/low levels of radiation exposure in aquatic environment. Fishes are very important as a group of ecologically and commercially important non-human biota and are often used as a bioindicators of aquatic pollution. Present work was carried out to assess the genotoxic effect of gamma radiation on fresh water fish Cyprinus carpio (common carp) in vivo using comet assay. Fishes were irradiated with 2, 4, 6, 8 and 10 Gy of gamma rays using a teletherapy machine and comet assay was performed on nucleated erythrocytes after 24, 48 and 72 h of irradiation . A significant increase in % tail DNA was observed at all the doses of gamma radiation as compared to controls indicating radiation induced DNA damage in a dose-dependent manner. Maximum % tail DNA was observed at 24 h which gradually declined till 72 h, in a time-dependent manner. This decrease in damage may indicate repair of the damaged DNA and or loss of heavily damaged cells, over a period of time. The study reveals that the comet assay may be used as a sensitive and rapid method to detect genotoxicity of gamma radiation and other environmental pollutants in sentinel species. (author)

  2. Neutron, gamma ray and post-irradiation thermal annealing effects on power semiconductor switches

    International Nuclear Information System (INIS)

    Schwarze, G.E.; Frasca, A.J.

    1994-01-01

    The effects of neutrons and gamma rays on the electrical and switching characteristics of power semiconductor switches must be known and understood by the designer of the power conditioning, control, and transmission subsystem of space nuclear power systems. The SP-100 radiation requirements at 25 m from the nuclear source are a neutron fluence of 10 13 n/cm 2 and a gamma dose of 0.5 Mrads. Experimental data showing the effects of neutrons and gamma rays on the performance characteristics of power-type NPN Bipolar Junction Transistors (BJTs), Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs), and Static Induction Transistors (SITs) are given in this paper. These three types of devices were tested at radiation levels which met or exceeded the SP-100 requirements. For the SP-100 radiation requirements, the BJTs were found to be most sensitive to neutrons, the MOSFETs were most sensitive to gamma rays, and the SITs were only slightly sensitive to neutrons. Post-irradiation thermal anneals at 300 K and up to 425 K were done on these devices and the effectiveness of these anneals are also discussed

  3. Neutron, gamma ray and post-irradiation thermal annealing effects on power semiconductor switches

    Science.gov (United States)

    Schwarze, G. E.; Frasca, A. J.

    1991-01-01

    The effects of neutron and gamma rays on the electrical and switching characteristics of power semiconductor switches must be known and understood by the designer of the power conditioning, control, and transmission subsystem of space nuclear power systems. The SP-100 radiation requirements at 25 m from the nuclear source are a neutron fluence of 10(exp 13) n/sq cm and a gamma dose of 0.5 Mrads. Experimental data showing the effects of neutrons and gamma rays on the performance characteristics of power-type NPN Bipolar Junction Transistors (BJTs), Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs), and Static Induction Transistors (SITs) are presented. These three types of devices were tested at radiation levels which met or exceeded the SP-100 requirements. For the SP-100 radiation requirements, the BJTs were found to be most sensitive to neutrons, the MOSFETs were most sensitive to gamma rays, and the SITs were only slightly sensitive to neutrons. Post-irradiation thermal anneals at 300 K and up to 425 K were done on these devices and the effectiveness of these anneals are also discussed.

  4. Use of borated polyethylene to improve low energy response of a prompt gamma based neutron dosimeter

    Energy Technology Data Exchange (ETDEWEB)

    Priyada, P.; Ashwini, U.; Sarkar, P.K., E-mail: pradip.sarkar@manipal.edu

    2016-05-21

    The feasibility of using a combined sample of borated polyethylene and normal polyethylene to estimate neutron ambient dose equivalent from measured prompt gamma emissions is investigated theoretically to demonstrate improvements in low energy neutron dose response compared to only polyethylene. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of boron, hydrogen and carbon prompt gamma emissions to mono energetic neutrons. The weighted least square method is employed to arrive at the best linear combination of these responses that approximates the ICRP fluence to dose conversion coefficients well in the energy range of 10{sup −8} MeV to 14 MeV. The configuration of the combined system is optimized through FLUKA simulations. The proposed method is validated theoretically with five different workplace neutron spectra with satisfactory outcome. - Highlights: • An improved method is proposed for estimating H⁎(10) using prompt gamma emissions. • A combination of BHDPE and HDPE cylinders is used as a sample. • Linear combination of prompt gamma intensities approximates ICRP-DCC closely. • Feasibility of the method was tested theoretically using workplace neutron spectra.

  5. Total body-calcium measurements: comparison of two delayed-gamma neutron activation facilities

    International Nuclear Information System (INIS)

    Ma, R.; Ellis, K.J.; Shypailo, R.J.; Pierson, R.N. Jr.

    1999-01-01

    This study compares two independently calibrated delayed-gamma neutron activation (DGNA) facilities, one at the Brookhaven National Laboratory (BNL), Upton, New York, and the other at the Children's Nutrition Research Center (CNRC), Houston, Texas that measure total body calcium (TBCa). A set of BNL phantoms was sent to CNRC for neutron activation analysis, and a set of CNRC phantoms was measured at BNL. Both facilities showed high precision (<2%), and the results were in good agreement, within 5%. (author)

  6. Neutron activation analysis for sulphur in coal samples and moisture content by gamma-ray transmission

    International Nuclear Information System (INIS)

    Selvi, S.

    1993-01-01

    A neutron activation analysis method is described for the determination of sulphur in coal samples by analysing the beta spectrum emitted from 32 P and 33 P following the reactions 32 S(n, p) 32 P and 33 S(n, p) 33 P using 252 Cf as a source of neutrons. The transmission of the combined gamma-rays emitted from three 137 Cs and three 241 Am sources is used to measure the water content of the coal samples. (author)

  7. Quantitative and qualitative applications of the neutron-gamma borehole logging

    International Nuclear Information System (INIS)

    Charbucinski, J.; Aylmer, J.A.; Eisler, P.L.; Borsaru, M.

    1989-01-01

    Two neutron-γ borehole logging applications are described. In a quantitative application of the prompt-gamma neutron-activation analysis (PGNAA) technique, research was carried out both in the laboratory and at a mine to establish a suitable borehole logging technology for manganese-grade predictions. As an example of the qualitative application of PGNAA, the use of this method has been demonstrated for the determination of lithology. (author)

  8. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  9. EJ-309 pulse shape discrimination performance with a high gamma-ray-to-neutron ratio and low threshold

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, A.C., E-mail: Alexis.C.Kaplan@gmail.com [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48104 (United States); Nuclear Engineering and Nonproliferation Division, Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, M.; Enqvist, A.; Dolan, J.L.; Pozzi, S.A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48104 (United States)

    2013-11-21

    Measuring neutrons in the presence of high gamma-ray fluence is a challenge with multi-particle detectors. Organic liquid scintillators such as the EJ-309 are capable of accurate pulse-shape discrimination (PSD) but the chance for particle misclassification is not negligible for some applications. By varying the distance from an EJ-309 scintillator to a strong-gamma-ray source and keeping a weak-neutron source at a fixed position, various gamma-to-neutron ratios can be measured and PSD performance can be quantified. Comparing neutron pulse-height distributions allows for pulse-height specific PSD evaluation, and quantification and visualization of deviation from {sup 252}Cf alone. Even with the addition of the misclassified gamma-rays, the PSD is effective in separating particles so that neutron count rate can be predicted with less than 10% error up to a gamma-to-neutron ratio of almost 650. For applications which can afford a reduction in neutron detection efficiency, PSD can be sufficiently effective in discriminating particles to measure a weak neutron source in a high gamma-ray background. -- Highlights: •We measure neutrons in a high photon background with EJ-309 liquid scintillators. •A low threshold is used to test the limits of particle discrimination. •A weak neutron signal is detectable with a gamma/neutron ratio as high as 770. •Photon pileup most commonly adds to error in classification of neutrons. •Neutron count rates are within 10% of expected rate under high gamma background.

  10. EJ-309 pulse shape discrimination performance with a high gamma-ray-to-neutron ratio and low threshold

    International Nuclear Information System (INIS)

    Kaplan, A.C.; Flaska, M.; Enqvist, A.; Dolan, J.L.; Pozzi, S.A.

    2013-01-01

    Measuring neutrons in the presence of high gamma-ray fluence is a challenge with multi-particle detectors. Organic liquid scintillators such as the EJ-309 are capable of accurate pulse-shape discrimination (PSD) but the chance for particle misclassification is not negligible for some applications. By varying the distance from an EJ-309 scintillator to a strong-gamma-ray source and keeping a weak-neutron source at a fixed position, various gamma-to-neutron ratios can be measured and PSD performance can be quantified. Comparing neutron pulse-height distributions allows for pulse-height specific PSD evaluation, and quantification and visualization of deviation from 252 Cf alone. Even with the addition of the misclassified gamma-rays, the PSD is effective in separating particles so that neutron count rate can be predicted with less than 10% error up to a gamma-to-neutron ratio of almost 650. For applications which can afford a reduction in neutron detection efficiency, PSD can be sufficiently effective in discriminating particles to measure a weak neutron source in a high gamma-ray background. -- Highlights: •We measure neutrons in a high photon background with EJ-309 liquid scintillators. •A low threshold is used to test the limits of particle discrimination. •A weak neutron signal is detectable with a gamma/neutron ratio as high as 770. •Photon pileup most commonly adds to error in classification of neutrons. •Neutron count rates are within 10% of expected rate under high gamma background

  11. The Fermi Gamma-Ray Space Telescope, Exploding Stars, Neutron Stars, and Black Holes

    Science.gov (United States)

    Thompson, David J.

    2010-01-01

    Since August, 2008, the Fermi Gamma-ray Space Telescope has been scanning the sky, producing a full-sky image every three hours. These cosmic gamma-rays come from extreme astrophysical phenomena, many related to exploding stars (supernovae) or what these explosions leave behind: supernova remnants, neutron stars, and black holes. This talk uses sample Fermi results, plus simple demonstrations, to illustrate the exotic properties of these endpoints of stellar evolution.

  12. Thermal neutron capture cross section for Fe-56(n,gamma)

    Czech Academy of Sciences Publication Activity Database

    Firestone, R. B.; Belgya, T.; Krtička, M.; Bečvář, F.; Szentmiklosi, L.; Tomandl, Ivo

    2017-01-01

    Roč. 95, č. 1 (2017), č. článku 014328. ISSN 2469-9985 R&D Projects: GA ČR GA13-07117S; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : neutron cross section * gamma gamma-coincidence data Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 3.820, year: 2016

  13. Measurements of angular and energy distributions of gamma-rays resulting from neutron interactions in shielding barriers

    International Nuclear Information System (INIS)

    Makarious, A.S.; Maayouf, R.M.A.; Megahid, R.

    1978-01-01

    Measurements of both angular and energy distributions of secondary gamma resulting from interactions of neutrons emerging from one of the ET-RR-1 reactor beam holes, in barriers from iron, lead and water are reported. The measurements were carried out, both with a bare neutron beam and with the beam being transmitted through a B4C. Filter, using a stilbene crystal gamma spectrometer. The spectrometer applies discrimination between neutrons and gammas according to the difference in decay times of the scintillations produced by them in stilbene. The described angular distributions resulted from measurements made at different angles of neutron incidence and with three different thicknesses of each sample

  14. Determination of dose components in mixed gamma neutron fields by use of high pressure ionization chambers

    International Nuclear Information System (INIS)

    Golnik, N.; Pliszczynski, T.; Wysocka, A.; Zielczynski, M.

    1985-01-01

    The two ionization chamber method for determination of dose components in mixed γ-neutron field has been improved by increasing gas pressure in the chambers up to some milions pascals. Advantages of high pressure gas filling are the followings: 1) significant reduction of the ratio of neutron-to gamma sensitivity for the hydrogen-free chamber, 2) possibility of sensitivity correction for both chambers by application of appropriate voltage, 3) high sensitivity for small detectors. High-pressure, pen-like ionization chambers have been examined in fields of different neutron sources: a TE-chamber, filled with 0.2 MPa of quasi-TE-gas and a conductive PTFE chamber, filled with 3.1 MPa of CO 2 . The ratio of neutron-to-gamma sensitivity for the PTFE chamber, operated at electrical field strength below 100 V/cm, has not exceeded 0.01 for neutrons with energy below 8 MeV. Formula is presented for calculation of this ratio for any high-pressure, CO 2 -filled ionization chamber. Contribution of gamma component to total tissue dose in the field of typical neutron sources has been found to be 3 to 70%

  15. Passive assay of plutonium metal plates using a fast-neutron multiplicity counter

    Energy Technology Data Exchange (ETDEWEB)

    Di Fulvio, A., E-mail: difulvio@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Shin, T.H.; Jordan, T.; Sosa, C.; Ruch, M.L.; Clarke, S.D. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Pozzi, S.A. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

    2017-05-21

    We developed a fast-neutron multiplicity counter based on organic scintillators (EJ-309 liquid and stilbene). The system detects correlated photon and neutron multiplets emitted by fission reactions, within a gate time of tens of nanoseconds. The system was used at Idaho National Laboratory to assay a variety of plutonium metal plates. A coincidence counting strategy was used to quantify the {sup 240}Pu effective mass of the samples. Coincident neutrons, detected within a 40-ns coincidence window, show a monotonic trend, increasing with the {sup 240}Pu-effective mass (in this work, we tested the 0.005–0.5 kg range). After calibration, the system estimated the {sup 240}Pu effective mass of an unknown sample ({sup 240}Pu{sub eff} >50 g) with an uncertainty lower than 1% in a 4-min assay time.

  16. Efficiency and attenuation correction factors determination in gamma spectrometric assay of bulk samples using self radiation

    International Nuclear Information System (INIS)

    Haddad, Kh.

    2009-02-01

    Gamma spectrometry forms the most important and capable tool for measuring radioactive materials. Determination of the efficiency and attenuation correction factors is the most tedious problem in the gamma spectrometric assay of bulk samples. A new experimental and easy method for these correction factors determination using self radiation was proposed in this work. An experimental study of the correlation between self attenuation correction factor and sample thickness and its practical application was also introduced. The work was performed on NORM and uranyl nitrate bulk sample. The results of proposed methods agreed with those of traditional ones.(author)

  17. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    International Nuclear Information System (INIS)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R.

    2008-01-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  18. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    Energy Technology Data Exchange (ETDEWEB)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R. [McMaster Univ., Medical Physics and Applied Radiation Sciences Unit, Hamilton, Ontario (Canada)

    2008-07-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  19. Verification of Gamma-ray Sensitivity for BF3 Neutron Detection System

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Cho, Jin Bok; Lyou, Seok Jean

    2016-01-01

    The BF3(Boron Tri-Fluorides) gas filled neutron detector(hereafter BF3 Detector) is commonly used for nuclear reactor’s startup channel due to its relatively high neutron efficiency and good discrimination against gamma-ray backgrounds. In order to measure how much this gamma-ray will affect on BF3 neutron detector performance in view of gamma noise discrimination, Multi-Channel Analyzer(MCA) is utilized for spectrum based signal analysis. The pre-test of BF3 Detector should be performed in an area where the ionization does not exceed 2.5 micro Gy/Hr(Ref.1). In this paper, the discrimination level (Voltage Unit) is verified by experimentally measurement if that discrimination level is acceptable within the criteria or not before installation. The maximum discrimination level, so called LLD, is determined by experimentally measurement. This BF3 Detector (LND20372) is insensitive under 540 micro Gy/Hr of gamma ray and 0.3V of LLD could cut off a background and gamma induced signal in a laboratory. MCA could be a convenient tool for spectrum analysis of signals that induced from gamma ray and a time saving tool rather than oscilloscope investigation due to its function to integrate all input signals at a sudden duration

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  1. PROMETHEE: An Alpha Low Level Waste Assay System Using Passive and Active Neutron Measurement Methods

    International Nuclear Information System (INIS)

    Passard, Christian; Mariani, Alain; Jallu, Fanny; Romeyer-Dherbey, Jacques; Recroix, Herve; Rodriguez, Michel; Loridon, Joel; Denis, Caroline; Toubon, Herve

    2002-01-01

    The development of a passive-active neutron assay system for alpha low level waste characterization at the French Atomic Energy Commission is discussed. Less than 50 Bq[α] (about 50 μg Pu) per gram of crude waste must be measured in 118-l 'European' drums in order to reach the requirements for incinerating wastes. Detection limits of about 0.12 mg of effective 239 Pu in total active neutron counting, and 0.08 mg of effective 239 Pu coincident active neutron counting, may currently be detected (empty cavity, measurement time of 15 min, neutron generator emission of 1.6 x 10 8 s -1 [4π]). The most limiting parameters in terms of performances are the matrix of the drum - its composition (H, Cl...), its density, and its heterogeneity degree - and the localization and self-shielding properties of the contaminant

  2. Method for measuring and evaluation dose equivalent rate from fast neutrons in mixed gamma-neutron fields around particles accelerators

    International Nuclear Information System (INIS)

    Cruceru, I.; Sandu, M.; Cruceru, M.

    1994-01-01

    A method for measuring and evaluation of doses and dose equivalent rate in mixed gamma- neutron fields is discussed in this paper. The method is basedon a double detector system consist of an ionization chamber with components made from a plastic scintillator, coupled to on photomultiplier. Generally the radiation fields around accelerators are complex, often consisting of many different ionizing radiations extending over a broad range of energies. This method solve two major difficulties: determination of response functions of radiation detectors; interpretation of measurement and determination of accuracy. The discrimination gamma-fast neutrons is assured directly without a pulse shape discrimination circuit. The method is applied to mixed fields in which particle energies are situated in the energy range under 20 MeV and an izotropic emision (Φ=10 4 -10 11 n.s -1 ). The dose equivalent rates explored is 0.01mSV--0.1SV

  3. Elemental analysis technique based on detecting gamma-rays from interactions of neutrons with medium

    International Nuclear Information System (INIS)

    Pospisil, S.; Janout, Z.; Vobecky, M.

    1979-01-01

    The methods are discussed of carbon content determination in large amounts of material by detecting 4438 keV gamma radiation accompanying inelastic scattering of neutrons from a radionuclide neutron source. Presented are the methodological analysis of the problem, the results of test measurements, and methodological recommendations for the practical application of the method. Test measurements were conducted on fly ash, limestone and brown coal in amounts of approximately 5 kg for each material sample, using an Am-Be neutron source. The determined sensitivity thresholds corresponded to the carbon concentration of 5 to 10% w.w. (S.P.)

  4. Interaction effect of gamma rays and thermal neutrons on the inactivation of odontoglossum ringspot virus isolated from orchid

    International Nuclear Information System (INIS)

    Mori, Itsuhiko; Inouye, Narinobu.

    1977-01-01

    The effect of gamma rays or thermal neutrons and their interaction effects on the inactivation of the infectivity of Odontoglossum ringspot virus (ORSV) in buffered crude sap of the plant tissue were studied. The inactivation effect of gamma ray on ORSV varied in different ionic strength of the phosphate buffer solutions. Borax enhanced this effect. In interaction effect of gamma and neutron irradiation, irradiation orders, that is, n → γ and γ → n, gave different inactivation pattern. (author)

  5. Neutron detection in a high gamma-ray background with EJ-301 and EJ-309 liquid scintillators

    International Nuclear Information System (INIS)

    Stevanato, L.; Cester, D.; Nebbia, G.; Viesti, G.

    2012-01-01

    Using a fast digitizer, the neutron–gamma discrimination capability of the new liquid scintillator EJ-309 is compared with that obtained using standard EJ-301. Moreover the capability of both the scintillation detectors to identify a weak neutron source in a high gamma-ray background is demonstrated. The probability of neutron detection is PD=95% at 95% confidence level for a gamma-ray background corresponding to a dose rate of 100 μSv/h.

  6. Determination of protein content in grains by radioactive thermal neutron capture prompt gamma rays analysis

    International Nuclear Information System (INIS)

    Carbonari, A.W.

    1983-01-01

    The radioactive thermal neutron capture prompt gamma rays technique can be used to determinate the nitrogen content in grains without chemical destruction, with good precision and relative rapidity. This determination is based on the detection of prompt gamma rays emitted by the 14 N(n,γ) 15 N reaction product. The samples has been irradiated the tanGencial tube of the IEA-R1 research reator and a pair spectrometer has been used for the detection of the prompt gamma rays. The nitrogen content is determinated in several samples of soybean, commonbean, peas and rice, and the results is compared with typical nitrogen content for each grain. (Autor) [pt

  7. Radiation hygiene aspects of mixed neutron-gamma field dosimetry

    International Nuclear Information System (INIS)

    Nikodemova, O.; Hrabovcova, A.

    1982-01-01

    Various possibilities are analyzed of determining the dose equivalent of neutrons, as is the reliability of the techniques and the correct interpretation for the purposes of radiation hygiene. (author)

  8. A new, passive dosemeter for gamma, beta and neutron radiations

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L A; Stokes, R P, E-mail: rpstokes@dstl.gov.uk [Defence Science and Technology Laboratory, Environmental Sciences Department, Alverstoke, Gosport, Hants, PO12 2DL (United Kingdom)

    2011-03-01

    The Defence Science and Technology Laboratory (Dstl) provides personal radiation dosimetry to the UK Ministry of Defence. Dstl has recently developed a dosemeter that is based on a combination of thermoluminescent and etched-track detectors. The Dstl Combined Dosemeter is capable of assessing doses due to photons, beta particles and neutrons. This paper presents the laboratory type testing results for the Combined Dosemeter, and also describes the procedure for calibrating the dosemeter for use in workplace neutron fields. The Combined Dosemeter meets the type test requirements that are relevant to its intended applications, and gives neutron doses that are within 50% of the true dose in the workplaces in which it is used, even when the wearer has the potential to be exposed to a variety of neutron spectra (e.g. on board nuclear-powered submarines).

  9. A new, passive dosemeter for gamma, beta and neutron radiations

    International Nuclear Information System (INIS)

    Jones, L A; Stokes, R P

    2011-01-01

    The Defence Science and Technology Laboratory (Dstl) provides personal radiation dosimetry to the UK Ministry of Defence. Dstl has recently developed a dosemeter that is based on a combination of thermoluminescent and etched-track detectors. The Dstl Combined Dosemeter is capable of assessing doses due to photons, beta particles and neutrons. This paper presents the laboratory type testing results for the Combined Dosemeter, and also describes the procedure for calibrating the dosemeter for use in workplace neutron fields. The Combined Dosemeter meets the type test requirements that are relevant to its intended applications, and gives neutron doses that are within 50% of the true dose in the workplaces in which it is used, even when the wearer has the potential to be exposed to a variety of neutron spectra (e.g. on board nuclear-powered submarines).

  10. Dose-response curve for blood exposed to gamma-neutron mixed field by conventional cytogenetic method

    International Nuclear Information System (INIS)

    Brandao, Jose Odinilson de C.; Souza, Priscilla L.G.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

    2009-01-01

    There is increasing concern about airline crew members (about one million worldwide) are exposed to measurable neutrons doses. Historically, cytogenetic biodosimetry assays have been based on quantifying asymmetrical chromosome alterations (dicentrics, centric rings and acentric fragments) in mytogen-stimulated T-lymphocytes in their first mitosis after radiation exposure. Increased levels of chromosome damage in peripheral blood lymphocytes are a sensitive indicator of radiation exposure and they are routinely exploited for assessing radiation absorbed dose after accidental or occupational exposure. Since radiological accidents are not common, not all nations feel that it is economically justified to maintain biodosimetry competence. However, dependable access to biological dosimetry capabilities is completely critical in event of an accident. In this paper the dose-response curve was measured for the induction of chromosomal alterations in peripheral blood lymphocytes after chronic exposure in vitro to neutron-gamma mixes field. Blood was obtained from one healthy donor and exposed to two neutron-gamma mixed field from sources 241 AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The evaluated absorbed doses were 0.2 Gy; 1.0 Gy and 2.5 Gy. The dicentric chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphase figures were analyzed for the presence of dicentrics by two experienced scorers after painted by giemsa 5%. Our preliminary results showed a linear dependence between radiations absorbed dose and dicentric chromosomes frequencies. Dose-response curve described in this paper will contribute to the construction of calibration curve that will be used in our laboratory for biological dosimetry. (author)

  11. Dose-response curve for blood exposed to gamma-neutron mixed field by conventional cytogenetic method

    Energy Technology Data Exchange (ETDEWEB)

    Brandao, Jose Odinilson de C.; Souza, Priscilla L.G.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F., E-mail: jodinilson@cnen.gov.b, E-mail: fflima@cnen.gov.b, E-mail: jasantos@cnen.gov.b [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Calixto, Merilane S.; Santos, Neide, E-mail: santos_neide@yahoo.com.b [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Genetica

    2009-07-01

    There is increasing concern about airline crew members (about one million worldwide) are exposed to measurable neutrons doses. Historically, cytogenetic biodosimetry assays have been based on quantifying asymmetrical chromosome alterations (dicentrics, centric rings and acentric fragments) in mytogen-stimulated T-lymphocytes in their first mitosis after radiation exposure. Increased levels of chromosome damage in peripheral blood lymphocytes are a sensitive indicator of radiation exposure and they are routinely exploited for assessing radiation absorbed dose after accidental or occupational exposure. Since radiological accidents are not common, not all nations feel that it is economically justified to maintain biodosimetry competence. However, dependable access to biological dosimetry capabilities is completely critical in event of an accident. In this paper the dose-response curve was measured for the induction of chromosomal alterations in peripheral blood lymphocytes after chronic exposure in vitro to neutron-gamma mixes field. Blood was obtained from one healthy donor and exposed to two neutron-gamma mixed field from sources {sup 241}AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The evaluated absorbed doses were 0.2 Gy; 1.0 Gy and 2.5 Gy. The dicentric chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphase figures were analyzed for the presence of dicentrics by two experienced scorers after painted by giemsa 5%. Our preliminary results showed a linear dependence between radiations absorbed dose and dicentric chromosomes frequencies. Dose-response curve described in this paper will contribute to the construction of calibration curve that will be used in our laboratory for biological dosimetry. (author)

  12. Fast neutron and gamma-ray transmission technique in mixed samples. MCNP calculations

    International Nuclear Information System (INIS)

    Perez, N.; Padron, I.

    2001-01-01

    In this paper the moisture in sand and also the sulfur content in toluene have been described by using the simultaneous fast neutron/gamma transmission technique (FNGT). Monte Carlo calculations show that it is possible to apply this technique with accelerator-based and isotopic neutron sources in the on-line analysis to perform the product quality control, specifically in the building materials industry and the petroleum one. It has been used particles from a 14MeV neutron generator and also from an Am-Be neutron source. The estimation of optimal system parameters like the efficiency, detection time, hazards and costs were performed in order to compare both neutron sources

  13. Self-absorption of neutron capture gamma-rays in gold samples

    International Nuclear Information System (INIS)

    Wisshak, K.; Walter, G.; Kaeppeler, F.

    1983-06-01

    The self absorption of neutron capture gamma rays in gold samples has been determined experimentally for two standard setups used in measurements of neutron capture cross sections. One makes use of an artificially collimated neutron beam and two C 6 D 6 detectors, the other of kinematically collimated neutrons and three Moxon-Rae detectors. Correction factors for an actual measurement of a neutron capture cross section using a gold standard of 1 mm thickness up to 12% were found for the first setup while they are only 4% for the second setup. The present data allow to determine the correction in an actual measurement with an accuracy of 0.5-1%. (orig.) [de

  14. Displacement damage caused by gamma-rays and neutrons on Au and Se.

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, Barney Lee [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-11-01

    This report documents theoretical calculations of displacement damage produced by gamma rays and neutrons on various materials. The average energy of the gamma rays was 1.24 MeV and 1.0 MeV for the neutrons. The fluence of the gamma rays was 1.2e14 γ/cm2 , for the neutrons it was 1.0e12 n/cm2. The initial materials of interest were Au and Se. The total doses of the gamma ray exposures were in the 100 kRad range for both elements. An equivalent electron fluence was approximated to be the same as the gamma ray fluence over one gamma ray attenuation length in both materials and at the same 1.24 MeV energy. The maximum recoil energy of the Au and Se for these electrons was calculated relativisticaly to be 29 and 72 eV respectively. The relativisitic McKinley and Feshbach theory for the atomic recoil cross sections produced by the electrons were in the 10s of mbarn range and an upper limit for the concentration of Frenkel pairs for the gamma ray exposures for both elements was in the ppb range. The Robinson Energy Partioning Theory for non-ionizing energy loss (NIEL) of ions in solids was used to calculate the concentration of Frenkel pairs produced by the 1 MeV neutrons, and this concentration was also in the ppb range for both Au and Se. Low damage levels like this can have effects on minority carrier recombination in semiconductors, but are not expected to have any effect on metals like Au, or metalloids such as Se.

  15. Effect of neutron and gamma radiations on zeolite and zeotype materials

    International Nuclear Information System (INIS)

    Durrani, S.K.

    1994-01-01

    The influence of gamma and (n, gamma)-radiation on the cation exchange and the structure of zeolite and zeotype materials has been studied. Samples were subjected to different doses of gamma-irradiation varying between 0.5 and 10 MGy and Neutron irradiation flux varied from 1.14 x 10/sup 17/ to 3.88 x /sup 10/sup 17/n cm/sup -2/. Structural effects consequent to gamma irradiation were examined by x-ray diffraction, electron scanning micrographs and FTIR measurements. Neutron and gamma-irradiation and not lead by any appreciable change in the structure, however, the displacement cations to locked-in sites results partial reduced barium and caesium uptake. The decrease of the intensities of the absorption bands of the hydroxy-groups reveals that gamma-irradiation has a strong dehydrating influence. THe effects of gamma-radiation on (UO/sub 2/)/sup 2+/ and Am/sup 3+/ uptake into NH/sub 4/-L and NH/sub 4/-SAPO-34 was also observed. K alpha of the uranyl ions increased with increasing pH up to 6.3. At pH > 3.5, the uranyl ions were precipitated and consequently K alpha values were continued to increased. (author)

  16. A maximum-likelihood reconstruction algorithm for tomographic gamma-ray nondestructive assay

    International Nuclear Information System (INIS)

    Prettyman, T.H.; Estep, R.J.; Cole, R.A.; Sheppard, G.A.

    1994-01-01

    A new tomographic reconstruction algorithm for nondestructive assay with high resolution gamma-ray spectroscopy (HRGS) is presented. The reconstruction problem is formulated using a maximum-likelihood approach in which the statistical structure of both the gross and continuum measurements used to determine the full-energy response in HRGS is precisely modeled. An accelerated expectation-maximization algorithm is used to determine the optimal solution. The algorithm is applied to safeguards and environmental assays of large samples (for example, 55-gal. drums) in which high continuum levels caused by Compton scattering are routinely encountered. Details of the implementation of the algorithm and a comparative study of the algorithm's performance are presented

  17. Advanced Laser-Compton Gamma-Ray Sources for Nuclear Materials Detection, Assay and Imaging

    Science.gov (United States)

    Barty, C. P. J.

    2015-10-01

    Highly-collimated, polarized, mono-energetic beams of tunable gamma-rays may be created via the optimized Compton scattering of pulsed lasers off of ultra-bright, relativistic electron beams. Above 2 MeV, the peak brilliance of such sources can exceed that of the world's largest synchrotrons by more than 15 orders of magnitude and can enable for the first time the efficient pursuit of nuclear science and applications with photon beams, i.e. Nuclear Photonics. Potential applications are numerous and include isotope-specific nuclear materials management, element-specific medical radiography and radiology, non-destructive, isotope-specific, material assay and imaging, precision spectroscopy of nuclear resonances and photon-induced fission. This review covers activities at the Lawrence Livermore National Laboratory related to the design and optimization of mono-energetic, laser-Compton gamma-ray systems and introduces isotope-specific nuclear materials detection and assay applications enabled by them.

  18. Fabrication of gamma sources using the neutron-gamma reactions of 238Pu13C

    International Nuclear Information System (INIS)

    Solinhac, I.; Maillard, C.; Donnet, L.

    2004-01-01

    A production campaign for 238 Pu 13 C sources with gamma fluence ranging from 2500 to 4500 gamma/s/4π at 6.13 MeV was carried out in 2002 in Atalante. An experimental study was undertaken to prepare the 238 PuC mixture, which is the most delicate step. This procedure is described together with the other steps in the source fabrication process: purification of a plutonium oxide batch, preparation of a nitric solution of 238 Pu, measurement of the gamma fluence of the PuC mixture before and after insertion into each of the two stainless steel capsules that constitute a PuN 2 O package, welding of the second envelope followed by leak testing, final measurement of the gamma fluence of the sealed source. This PuC sources fabrication procedure is effective: all the sources include the required gamma activity with an uncertainty on the gamma fluence of less than 10%. (authors)

  19. Gamma ray NDA assay system for total plutonium and isotopics in plutonium product solutions

    International Nuclear Information System (INIS)

    Cowder, L.R.; Hsue, S.T.; Johnson, S.S.; Parker, J.L.; Russo, P.A.; Sprinkle, J.K.; Asakura, Y.; Fukuda, T.; Kondo, I.

    1979-01-01

    A LASL-designed gamma-ray NDA instrument for assay of total plutonium and isotopics of product solutions at Tokai-Mura is currently installed and operating. The instrument is, optimally, a densitometer that uses radioisotopic sources for total plutonium measurements at the K absorption edge. The measured transmissions of additional gamma-ray lines from the same radioisotopic sources are used to correct for self-attenuation of passive gamma rays from plutonium. The corrected passive data give the plutonium isotopic content of freshly separated to moderately aged solutions. This off-line instrument is fully automated under computer control, with the exception of sample positioning, and operates routinely in a mode designed for measurement control. A one-half percent precision in total plutonium concentration is achieved with a 15-minute measurement

  20. Geiger-Mueller counter for mixed neutron-gamma beam dosimetry

    International Nuclear Information System (INIS)

    McDonald, J.C.; Ma, I.-C.

    1978-01-01

    A Geiger-Mueller (G-M) dosimeter has been constructed and employed to measure the gamma-ray component of absorbed dose in a cyclotron produced fast neutron field. This instrument is waterproof for measurements in a liquid medium, and read-out is accompanied with any standard scaler. (Auth.)

  1. Results on neutron and gamma-ray irradiation of electrolytic tiltmeters

    International Nuclear Information System (INIS)

    Calderon, A.; Calvo, E.; Figueroa, C.F.; Martinez-Rivero, C.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A.L.; Arce, P.; Barcala, J.M.; Ferrando, A.; Fuentes, J.; Josa, M.I.; Luque, J.M.; Molinero, A.; Navarrete, J.; Oller, J.C.; Valdivieso, P.; Fenyvesi, A.; Molnar, J.

    2004-01-01

    We report on irradiation studies done to a sample of high-precision electrolytic tiltmeters with gamma-rays, up to a maximum dose of 150 kGy, and neutrons, up to a maximum fluence of 1.5x10 14 cm -2 . The effect of the irradiation on their performance is discussed

  2. Up-dating of the RA-0 reactor shielding. Gamma and neutron isodoses

    International Nuclear Information System (INIS)

    Murua, Carlos A.; Chautemps, Norma A.; Ackerley, Alejandro F.; Alexeiew, Vladimiro

    1999-01-01

    A comparative analysis of the historical shielding configurations of the RA-0 reactor is performed and the comparison methodology is described. The gamma and neutron dose mapping of the last two stages of the reactor shielding has been carried out and the results are analysed

  3. Measurements of Soil Carbon by Neutron-Gamma Analysis in Static and Scanning Modes.

    Science.gov (United States)

    Yakubova, Galina; Kavetskiy, Aleksandr; Prior, Stephen A; Torbert, H Allen

    2017-08-24

    The herein described application of the inelastic neutron scattering (INS) method for soil carbon analysis is based on the registration and analysis of gamma rays created when neutrons interact with soil elements. The main parts of the INS system are a pulsed neutron generator, NaI(Tl) gamma detectors, split electronics to separate gamma spectra due to INS and thermo-neutron capture (TNC) processes, and software for gamma spectra acquisition and data processing. This method has several advantages over other methods in that it is a non-destructive in situ method that measures the average carbon content in large soil volumes, is negligibly impacted by local sharp changes in soil carbon, and can be used in stationary or scanning modes. The result of the INS method is the carbon content from a site with a footprint of ~2.5 - 3 m 2 in the stationary regime, or the average carbon content of the traversed area in the scanning regime. The measurement range of the current INS system is >1.5 carbon weight % (standard deviation ± 0.3 w%) in the upper 10 cm soil layer for a 1 hmeasurement.

  4. Dosimetry of the Embalse nuclear power plant neutron/gamma mixed fields

    International Nuclear Information System (INIS)

    Salas, C.A.

    1990-01-01

    The aim of this work is to describe the method used at the Embalse nuclear power plant for carrying out personal dosimetry of the agents affected to the tasks on the Embalse nuclear power plant neutron-gamma mixed fields. (Author) [es

  5. Evaluation of a gamma-spectroscopy gauge for uranium-plutonium assay

    International Nuclear Information System (INIS)

    Notea, A.; Segal, Y.

    1976-01-01

    A procedure is presented for the characterization of a gamma passive method for non-destructive analysis of nuclear fuel. The approachh provides an organized and systematic way for optimizing the assay system. The key function is the relative resolving power defined as the smallest relative change in the quantity of radionuclide measured that may be detected within a certain confidence level. This function is derived for nuclear fuel employing a model based on empirical parameters. The ability to detect changes in fuels of binary and trinary compositions with a 50-cm 3 Ge(Li) at a 1-min counting period is discussed. As an example to a binary composition, an enriched uranium fuel was considered. The 185-keV and 1001-keV gamma lines are used for the assay of 235 U and 238 U, respectively. As a trinary composition a plutonium-containing fuel was examined. The plutonium was identified by the 414-keV gamma line. The interference of the high-energy lines is carefully analysed, and numerical results are presented. For both cases the range of measurement under specific accuracy demands is determined. The approac described is suitable also for evaluation of other passive as well as active assay methods. (author)

  6. Directional Stand-off Detection of Fast Neutrons and Gammas Using Angular Scattering Distributions

    Energy Technology Data Exchange (ETDEWEB)

    Vanier P. e.; Dioszegi, I.; Salwen, C.; Forman, L.

    2009-10-25

    We have investigated the response of a DoubleScatter Neutron Spectrometer (DSNS) for sources at long distances (gr than 200 meters). We find that an alternative method for analyzing double scatter data avoids some uncertainties introduced by amplitude measurements in plastic scintillators.Time of flight is used to discriminate between gamma and neutron events, and the kinematic distributions of scattering angles are assumed to apply. Non-relativistic neutrons are most likely to scatter at 45°, while gammas with energies greater than 2 MeV are most likely to be forward scattered. The distribution of scattering angles of fission neutrons arriving from a distant point source generates a 45° cone, which can be back-projected to give the source direction. At the same time, the distribution of Compton-scattered gammas has a maximum in the forward direction, and can be made narrower by selecting events that deposit minimal energy in the first scattering event. We have further determined that the shape of spontaneous fission neutron spectra at ranges gr than 110 m is still significantly different from thecosmic ray background.

  7. Characteristic Investigation of Unfolded Neutron Spectra with Different Priori Information and Gamma Radiation Interference

    International Nuclear Information System (INIS)

    Kim, Bong Hwan

    2006-01-01

    Neutron field spectrometry using multi spheres such as Bonner Spheres (BS) has been almost essential in radiation protection dosimetry for a long time at workplace in spite of poor energy resolution because it is not asking the fine energy resolution but requiring easy operation and measurement performance over a wide range of energy interested. KAERI has developed and used extended BS system based on a LiI(Eu) scintillator as the representative neutron spectrometry system for workplace monitoring as well as for the quantification of neutron calibration fields such as those recommended by ISO 8529. Major topics in using BS are how close the unfolded spectra is the real one and to minimize the interference of gamma radiation in neutron/gamma mixed fields in case of active instrument such as a BS with a LiI(Eu) scintillator. The former is related with choosing a priori information when unfolding the measured data and the latter is depend on how to discriminate it in intense gamma radiation fields. Influence of a priori information in unfolding and effect of counting loss due to pile-up of signals for the KAERI BS system were investigated analyzing the spectral measurement results of Scattered Neutron Calibration Fields (SNCF)

  8. Neutron/gamma dose separation by the multiple-ion-chamber technique

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1983-01-01

    Many mixed n/γ dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and 60 Co gamma ray irradiation in air. The MORSE-CG coupled n/γ three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references

  9. Formation properties from high resolution neutron activation gamma-ray spectra

    International Nuclear Information System (INIS)

    Mellor, D.W.; Underwood, M.C.

    1985-01-01

    A neutron activation logging tool has been developed comprising a Five Curie /sup 241/ Am-Be neutron source and a large n-type hyper-pure germanium gamma-ray detector. The tool maintains a constant temperature cryogenic environment for periods in excess of twenty hours. No liquid nitrogen or other consumable material is used in the operating or recharging stages. A large calibration tank in simulated well-bore geometry has been constructed with sand bodies saturated with oil and low salinity water (14,000 ppm NaCl). In the water zone prompt neutron capture gamma-rays from silicon, hydrogen and chlorine were prominent; gamma-rays from inelastic scattering on oxygen and silicon were detected. No gamma-rays arising from inelastic scattering on carbon were detected. These data have been interpreted to yield the porosity, fluid saturations, salinity and matrix composition. In the oil zone, gamma-rays arising from inelastic scattering on oxygen, silicon and carbon were detected. The intensity of the carbon line was very poor, and inadequate for quantitative purposes

  10. Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

    Science.gov (United States)

    Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

    2018-03-01

    Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

  11. The effect of pulse pile-up on discrimination between neutrons and gamma rays

    International Nuclear Information System (INIS)

    Whittlestone, S.

    1980-01-01

    Pulse pile-up lengthens the rise-time of pulses. With an organic scintillator such as NE 213, pile-up can cause a short rise-time pulse originating from gamma rays to be interpreted by a rise-time analyser as a neutron. The degradation of pulse shape analyser performance at high count rates is shown to be directly related to pulse pile-up. Using this relationship, the contribution of piled-up gamma rays and neutrons to count rate related errors is calculated for a time-dependent fast neutron energy spectrum measurement. Errors of a few per cent occur even when the probability of a count per burst is as low as 0.01. (orig.)

  12. Distributions of neutron and gamma doses in phantom under a mixed field

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.

    1982-06-01

    A calculation program, based on Monte Carlo method, allowed to estimate the absorbed doses relatives to the reactor primary radiation, in a water cubic phantom and in cylindrical phantoms modelized from tissue compositions. This calculation is a theoretical approach of gamma and neutron dose gradient study in an animal phantom. PIN junction dosimetric characteristics have been studied experimentally. Air and water phantom radiation doses measured by PIN junction and lithium 7 fluoride, in reactor field have been compared to doses given by dosimetry classical techniques as tissue equivalent plastic and aluminium ionization chambers. Dosimeter responses have been employed to evaluate neutron and gamma doses in plastinaut (tissue equivalent plastic) and animal (piglet). Dose repartition in the piglet bone medulla has been also determined. This work has been completed by comparisons with Doerschell, Dousset and Brown results and by neutron dose calculations; the dose distribution related to lineic energy transfer in Auxier phantom has been also calculated [fr

  13. Gamma and neutron detection modeling in the nuclear detection figure of merit (NDFOM) portal

    International Nuclear Information System (INIS)

    Stroud, Phillip D.; Saeger, Kevin J.

    2009-01-01

    The Nuclear Detection Figure Of Merit (NDFOM) portal is a database of objects and algorithms for evaluating the performance of radiation detectors to detect nuclear material. This paper describes the algorithms used to model the physics and mathematics of radiation detection. As a first-principles end-to-end analysis system, it starts with the representation of the gamma and neutron spectral fluxes, which are computed with the particle and radiation transport code MCNPX. The gamma spectra emitted by uranium, plutonium, and several other materials of interest are described. The impact of shielding and other intervening material is computed by the method of build-up factors. The interaction of radiation with the detector material is computed by a detector response function approach. The construction of detector response function matrices based on MCNPX simulation runs is described in detail. Neutron fluxes are represented in a three group formulation to treat differences in detector sensitivities to thermal, epithermal, and fast neutrons.

  14. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  15. Evaluation of neutron and gamma-ray-production cross-section data for lead

    International Nuclear Information System (INIS)

    Fu, C.Y.; Perey, F.G.

    1975-01-01

    A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV have been prepared. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser--Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture. 15 tables, 36 figures, 104 references

  16. Investigation of gamma-ray sensitivity of neutron detectors based on thin converter films

    Energy Technology Data Exchange (ETDEWEB)

    Khaplanov, A; Hall-Wilton, R [European Spallation Source, P.O Box 176, SE-22100 Lund (Sweden); Piscitelli, F; Buffet, J-C; Clergeau, J-F; Correa, J; Esch, P van; Ferraton, M; Guerard, B [Institute Laue Langevin, Rue Jules Horowitz, FR-38042 Grenoble (France)

    2013-10-15

    Currently, many detector technologies for thermal neutron detection are in development in order to lower the demand for the rare {sup 3}He gas. Gas detectors with solid thin film neutron converters readout by gas proportional counter method have been proposed as an appropriate choice for applications where large area coverage is necessary. In this paper, we investigate the probability for {gamma}-rays to generate a false count in a neutron measurement. Simulated results are compared to measurement with {sup 10}B thin film prototypes and a {sup 3}He detector. It is demonstrated that equal {gamma}-ray rejection to that of {sup 3}He tubes is achieved with the new technology. The arguments and results presented here are also applicable to gas detectors with converters other than solid {sup 10}B layers, such as {sup 6}Li layers and {sup 10}BF{sub 3} gas.

  17. Gamma-ray emission spectra from spheres with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

    1989-01-01

    Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

  18. Gamma signatures of the C-BORD Tagged Neutron Inspection System

    Directory of Open Access Journals (Sweden)

    Sardet A.

    2018-01-01

    Full Text Available In the frame of C-BORD project (H2020 program of the EU, a Rapidly relocatable Tagged Neutron Inspection System (RRTNIS is being developed to non-intrusively detect explosives, chemical threats, and other illicit goods in cargo containers. Material identification is performed through gamma spectroscopy, using twenty NaI detectors and four LaBr3 detectors, to determine the different elements composing the inspected item from their specific gamma signatures induced by fast neutrons. This is performed using an unfolding algorithm to decompose the energy spectrum of a suspect item, selected by X-ray radiography and on which the RRTNIS inspection is focused, on a database of pure element gamma signatures. This paper reports on simulated signatures for the NaI and LaBr3 detectors, constructed using the MCNP6 code. First experimental spectra of a few elements of interest are also presented.

  19. Thermal neutron detector and gamma-ray spectrometer utilizing a single material

    Science.gov (United States)

    Stowe, Ashley; Burger, Arnold; Lukosi, Eric

    2017-05-02

    A combined thermal neutron detector and gamma-ray spectrometer system, including: a detection medium including a lithium chalcopyrite crystal operable for detecting thermal neutrons in a semiconductor mode and gamma-rays in a scintillator mode; and a photodetector coupled to the detection medium also operable for detecting the gamma rays. Optionally, the detection medium includes a .sup.6LiInSe.sub.2 crystal. Optionally, the detection medium comprises a compound formed by the process of: melting a Group III element; adding a Group I element to the melted Group III element at a rate that allows the Group I and Group III elements to react thereby providing a single phase I-III compound; and adding a Group VI element to the single phase I-III compound and heating; wherein the Group I element includes lithium.

  20. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Mozin, Vladimir [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Campbell, Luke [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hunt, Alan W. [Iowa State Univ., Ames, IA (United States); Reedy, Edward T.E. [Iowa State Univ., Ames, IA (United States); Seipel, Heather A. [Iowa State Univ., Ames, IA (United States)

    2015-09-28

    This project has been a collaborative effort of researchers from four National Laboratories, Lawrence Berkley National Laboratory (LBNL), Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL), and Idaho State University’s (ISU) Idaho Accelerator Center (IAC). Experimental measurements at the Oregon State University (OSU) were also supported. The research included two key components, a strong experimental campaign to characterize the delayed gamma-ray signatures of the isotopes of interests and of combined targets, and a closely linked modeling effort to assess system designs and applications. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Detailed signature knowledge is essential for analyzing the capabilities of the delayed gamma technique, optimizing measurement parameters, and specifying neutron source and gamma-ray detection system requirements. The research was divided into three tasks: experimental measurements, characterization of fission yields, and development of analysis methods (task 1), modeling in support of experiment design and analysis and for the assessment of applications (task 2), and high-rate gamma-ray detector studies (task 3).

  1. Angular resolution study of a combined gamma-neutron coded aperture imager for standoff detection

    International Nuclear Information System (INIS)

    Ayaz-Maierhafer, Birsen; Hayward, Jason P.; Ziock, Klaus P.; Blackston, Matthew A.; Fabris, Lorenzo

    2013-01-01

    Nuclear threat source observables at standoff distances of tens of meters from mCi class sources include both gamma-rays and neutrons. This work uses simulations to investigate the effects of the angular resolution of a mobile gamma-ray and neutron coded aperture imaging system upon orphan source detection significance and specificity. The design requires maintaining high sensitivity and specificity while keeping the system size as compact as possible to reduce weight, footprint, and cost. A mixture of inorganic and organic scintillators was considered in the detector plane for high sensitivity to both gamma-rays and fast neutrons. For gamma-rays (100 to 2500 keV) and fission spectrum neutrons, angular resolutions of 1–9° and radiation angles of incidence appropriate for mobile search were evaluated. Detection significance for gamma-rays considers those events that contribute to the photopeak of the image pixel corresponding the orphan source location. For detection of fission spectrum neutrons, energy depositions above a set pulse shape discrimination threshold were tallied. The results show that the expected detection significance for the system at an angular resolution of 1° is significantly lower compared to its detection significance an angular resolution of ∼3–4°. An angular resolution of ∼3–4° is recommended both for better detection significance and improved false alarm rate, considering that finer angular resolution does not result in improved background rejection when the coded aperture method is used. Instead, over-pixelating the search space may result in an unacceptably high false alarm rate

  2. Comparison of the radiobiological effects of Boron neutron capture therapy (BNCT) and conventional Gamma Radiation

    International Nuclear Information System (INIS)

    Dagrosa, Maria A.; Carpano, Marina; Perona, Marina; Thomasz, Lisa; Juvenal, Guillermo J.; Pisarev, Mario; Pozzi, Emiliano; Thorp, Silvia

    2009-01-01

    BNCT is an experimental radiotherapeutic modality that uses the capacity of the isotope 10 B to capture thermal neutrons leading to the production of 4 He and 7 Li, particles with high linear energy transfer (LET). The aim was to evaluate and compare in vitro the mechanisms of response to the radiation arising of BNCT and conventional gamma therapy. We measured the survival cell fraction as a function of the total physical dose and analyzed the expression of p27/Kip1 and p53 by Western blotting in cells of colon cancer (ARO81-1). Exponentially growing cells were distributed into the following groups: 1) BPA (10 ppm 10 B) + neutrons; 2) BOPP (10 ppm 10 B) + neutrons; 3) neutrons alone; 4) gamma-rays. A control group without irradiation for each treatment was added. The cells were irradiated in the thermal neutron beam of the RA-3 (flux= 7.5 10 9 n/cm 2 sec) or with 60 Co (1Gy/min) during different times in order to obtain total physical dose between 1-5 Gy (±10 %). A decrease in the survival fraction as a function of the physical dose was observed for all the treatments. We also observed that neutrons and neutrons + BOPP did not differ significantly and that BPA was the more effective compound. Protein extracts of irradiated cells (3Gy) were isolated to 24 h and 48 h post radiation exposure. The irradiation with neutrons in presence of 10 BPA or 10 BOPP produced an increase of p53 at 24 h maintain until 48 h. On the contrary, in the groups irradiated with neutrons alone or gamma the peak was observed at 48 hr. The level of expression of p27/Kip1 showed a reduction of this protein in all the groups irradiated with neutrons (neutrons alone or neutrons plus boron compound), being more marked at 24 h. These preliminary results suggest different radiobiological response for high and low let radiation. Future studies will permit establish the role of cell cycle in the tumor radio sensibility to BNCT. (author)

  3. Radiation hardness of GaAs sensors against gamma-rays, neutrons and electrons

    Energy Technology Data Exchange (ETDEWEB)

    Šagátová, Andrea, E-mail: andrea.sagatova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Ilkovičova 3, 812 19 Bratislava (Slovakia); University Centre of Electron Accelerators, Slovak Medical University, Ku kyselke 497, 911 06 Trenčín (Slovakia); Zaťko, Bohumír; Dubecký, František [Institute of Electrical Engineering, Slovak Academy of Sciences, Dúbravská cesta 9, 841 04 Bratislava (Slovakia); Ly Anh, Tu [Faculty of Applied Science, University of Technology VNU HCM, 268 Ly Thuong Kiet Street, District 10, Ho Chi Minh City (Viet Nam); Nečas, Vladimír; Sedlačková, Katarína; Pavlovič, Márius [Institute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Ilkovičova 3, 812 19 Bratislava (Slovakia); Fülöp, Marko [University Centre of Electron Accelerators, Slovak Medical University, Ku kyselke 497, 911 06 Trenčín (Slovakia)

    2017-02-15

    Highlights: • Radiation hardness of SI GaAs detectors against gamma-rays, neutrons and electrons was compared. • Good agreement was achieved between the experimental results and displacement damage factor of different types of radiation. • CCE and FWHM first slightly improved (by 1–8%) and just then degraded with the cumulative dose. • An increase of detection efficiency with cumulative dose was observed. - Abstract: Radiation hardness of semi-insulating GaAs detectors against {sup 60}Co gamma-rays, fast neutrons and 5 MeV electrons was compared. Slight improvements in charge collection efficiency (CCE) and energy resolution in FWHM (Full Width at Half Maximum) were observed at low doses with all kinds of radiation followed by their degradation. The effect occurred at a dose of about 10 Gy of neutrons (CCE improved by 1%, FWHM by 5% on average), at 1 kGy of electrons (FWHM decreased by 3% on average) and at 10 kGy of gamma-rays (CCE raised by 5% and FWHM dropped by 8% on average), which is in agreement with the relative displacement damage of the used types of radiation. Gamma-rays of MeV energies are 1000-times less damaging than similar neutrons and electrons about 10-times more damaging than photons. On irradiating the detectors with neutrons and electrons, we observed a global increase in their detection efficiency, which was caused probably by enlargement of the active detector area as a consequence of created radiation defects in the base material. Detectors were still functional after a dose of 1140 kGy of ∼1 MeV photons, 104 kGy of 5 MeV electrons but only up to 0.576 kGy of fast (∼2 to 30 MeV) neutrons.

  4. Characterisation of neutron beam and gamma spectrometer for PGAA

    International Nuclear Information System (INIS)

    Revay, Zs.; Molnar, G.L.

    2001-01-01

    In the second project year great efforts have been devoted in Budapest to the development of methods and procedures for neutron beam characterisation and spectrometer calibration. These are described here to provide recipes for other laboratories. Some illustrative results obtained on the former thermal guide, and partly on the new cold neutron guide are also given. Preliminary results from the benchmark experiments on flux monitors titanium standard and an unknown sample are also reported. New k o factors for elements of highest priority will be measured on the cold beam only in the near future. (author)

  5. Evaluation of Kalman filters and genetic algorithms for delayed-neutron nondestructive assay data analyses

    International Nuclear Information System (INIS)

    Aumeier, S.E.; Forsmann, J.H.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile/fertile nuclei in various matrices is important in several areas of nuclear applications, including international and domestic safeguards, radioactive waste characterization, and nuclear facility operations. An analysis was performed to determine the feasibility of identifying the masses of individual fissionable isotopes from a cumulative delayed-neutron signal resulting form the neutron irradiation of several uranium and plutonium isotopes. The feasibility of two separate data-processing techniques was studied: Kalman filtering and genetic algorithms. The basis of each technique is reviewed, and the structure of the algorithms as applied to the delayed-neutron analysis problem is presented. The results of parametric studies performed using several variants of the algorithms are presented. The effect of including additional constraining information such as additional measurements and known relative isotopic concentration is discussed. The parametric studies were conducted using simulated delayed-neutron data representative of the cumulative delayed-neutron response following irradiation of a sample containing 238 U, 235 U, 239 Pu, and 240 Pu. The results show that by processing delayed-neutron data representative of two significantly different fissile/fertile fission ratios, both Kalman filters and genetic algorithms are capable of yielding reasonably accurate estimates of the mass of individual isotopes contained in a given assay sample

  6. Design of an electron-accelerator-driven compact neutron source for non-destructive assay

    Science.gov (United States)

    Murata, A.; Ikeda, S.; Hayashizaki, N.

    2017-09-01

    The threat of nuclear and radiological terrorism remains one of the greatest challenges to international security, and the threat is constantly evolving. In order to prevent nuclear terrorism, it is important to avoid unlawful import of nuclear materials, such as uranium and plutonium. Development of technologies for non-destructive measurement, detection and recognition of nuclear materials is essential for control at national borders. At Tokyo Institute of Technology, a compact neutron source system driven by an electron-accelerator has been designed for non-destructive assay (NDA). This system is composed of a combination of an S-band (2.856 GHz) RF-gun, a tungsten target to produce photons by bremsstrahlung, a beryllium target, which is suitable for use in generating neutrons because of the low threshold energy of photonuclear reactions, and a moderator to thermalize the fast neutrons. The advantage of this system can accelerate a short pulse beam with a pulse width less than 1 μs which is difficult to produce by neutron generators. The amounts of photons and neutron produced by electron beams were simulated using the Monte Carlo simulation code PHITS 2.82. When the RF-gun is operated with an average electron beam current of 0.1 mA, it is expected that the neutron intensities are 1.19 × 109 n/s and 9.94 × 109 n/s for incident electron beam energies of 5 MeV and 10 MeV, respectively.

  7. Trends in X-, gamma and neutron radiographic imaging at IGCAR Kalpakkam

    International Nuclear Information System (INIS)

    Venkatraman, B.; Raghu, N.; Menaka, M.; Anandraj, R.

    2015-01-01

    In the nuclear fuel cycle, right from raw material stage through fabrication and in service inspection upto the retirement of the component, NDE is an indispensable tool. While X- and gamma radiography is quite common, neutron radiography is a very efficient and complementary tool which can enhance investigations in the field of non-destructive testing as well as in many fundamental research applications. The main advantage of neutrons compared to X-rays is its ability to penetrate heavy elements and also image light elements (i.e. with low atomic numbers) such as hydrogen, water, carbon etc. This is because, neutrons interact with the nucleus rather than with the outer electron in the shell. This also makes it possible to distinguish between different isotopes of the same element by neutron radiography. The KAMINI reactor at IGCAR is a versatile and unique facility wherein extensive work has been undertaken on neutron radiography and activation analysis. Apart from conventional neutron radiography using transfer technique, real time neutron imaging of fuel pins and other objects have also been carried out. Using Beam purity indicator and sensitivity indicator, the neutron beam from KAMINI has also been characterized. This paper focuses on the developments and applications of digital imaging NDE using X-, gamma and neutrons at IGCAR. Both 2-dimensional imaging and -D tomography has been undertaken. Case studies undertaken for strategic and core industries including societal applications such as in cultural heritage is also highlighted. Advanced image processing and analysis has also been applied for enhancing the sensitivity and better defect quantification

  8. Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

    International Nuclear Information System (INIS)

    Spyrou, N.M.; Awotwi-Pratt, J.B.; Williams, A.M.

    2004-01-01

    An activation analysis facility based on an isotopic neutron source (185 GBq 241 Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, it was attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. (author)

  9. Gamma-ray production cross sections for MeV neutrons

    International Nuclear Information System (INIS)

    Kitazawa, Hideo; Harima, Yoshiko; Yamakoshi, Hisao; Sano, Yuji; Kobayashi, Tsuguyuki.

    1979-01-01

    Gamma-ray production cross section and spectra for 1- to 20-MeV neutrons were theoretically obtained, which were requested for heating calculations, for shielding design calculations, and for material damage estimates. Calculations were carried out for Al, Si, Ca, Fe, Ni, Cu, Nb, Ta, Au, and Pb, using a spin-dependent evaporation model without the parity conservation and including the dipole and quardupole gamma-ray transitions. The results were compared with the experimental data measured in ORNL to confirm the availability of this model in applications. In addition, the effects on the gamma-ray production cross section of the optical potential, level density, yrast level, and radiation width were investigated in detail. The conclusions are: 1) the use of the optical potential which gives the correct total reaction cross section is essential to gamma-ray production calculations, 2) the gamma-ray production cross section is not so sensitive to the choice of level density parameters, 3) the inclusion of yrast levels is necessary in dealing with the competition of the neutron and gamma-ray emissions from highly excited states, and 4) the Brink-Axel type's radiation width is unsuitable to be applied to radiative capture processes. (author)

  10. Nuclear structure studies on 178Hf by means of neutron induced gamma and electron spectroscopy

    International Nuclear Information System (INIS)

    Al Mamun Imtiazul Haque.

    1985-01-01

    By means of thermal and epithermal neutron captures the nucleus 178 Hf was studied. With high-resolution spectrometers the gamma transitions and conversion electrons were measured. By the found energies, intensities, and multipolarities the level scheme of 178 Hf could be essentially improved and extended. Totally 270 secondary (from 600 gamma lines) and 39 primary gamma transitions were used in order to establish the level scheme with 66 levels in 18 rotational bands. For this 92% of all gamma intensities were used. Several new rotational bands were established. By improved gamma energies the level scheme below 2 MeV for spins between 0 and 6 is well confirmed. Moreover by the resolution of several multiplets the decay structure of the levels could be explained. The thermal neutron capture state results from the primary gamma transitions to Q n =7626.34 (23) keV. Electrical monopole transitions from several states were studied in order to determine the X(E0/E2) values. (orig./HSI) [de

  11. Measurement of prompt fission gamma-ray spectra in fast neutron-induced fission

    International Nuclear Information System (INIS)

    Laborie, J.M.; Belier, G.; Taieb, J.

    2012-01-01

    Knowledge of prompt fission gamma-ray emission has been of major interest in reactor physics for a few years. Since very few experimental spectra were ever published until now, new measurements would be also valuable to improve our understanding of the fission process. An experimental method is currently being developed to measure the prompt fission gamma-ray spectrum from some tens keV up to 10 MeV at least. The mean multiplicity and total energy could be deduced. In this method, the gamma-rays are measured with a bismuth germanate (BGO) detector which has the advantage to present a high P/T ratio and a high efficiency compared to other gamma-ray detectors. The prompt fission neutrons are rejected by the time of flight technique between the BGO detector and a fission trigger given by a fission chamber or a scintillating active target. Energy and efficiency calibration of the BGO detector were carried out up to 10.76 MeV by means of the Al-27(p, gamma) reaction. First prompt fission gamma-ray spectrum measurements performed for the spontaneous fission of Cf-252 and for 1.7 and 15.6 MeV neutron-induced fission of U-238 at the CEA, DAM, DIF Van de Graaff accelerator, will be presented. (authors)

  12. 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

  13. Neutron and gamma radiography with a research reactor

    International Nuclear Information System (INIS)

    Reijonen, Heikki

    1973-01-01

    The thesis consists of eight publications, which describe the instrumental analysis of both neutron and γ-radiography. The emphasis is on various research applications, especially solidification and segregation in Sn-Cd and Sn-In alloys. (L.M.K.)

  14. Proton Neutron Gamma-X Detection (PNGXD): An introduction to contrast agent detection during proton therapy via prompt gamma neutron activation

    Science.gov (United States)

    Gräfe, James L.

    2017-09-01

    experimental work are required to determine the feasibility of this new technique termed Proton Neutron Gamma-X Detection (PNGXD). The initial concept of this procedure is presented in this paper as well as future research directions.

  15. Further Evaluation of the Neutron Resonance Transmission Analysis (NRTA) Technique for Assaying Plutonium in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    J. W. Sterbentz; D. L. Chichester

    2011-09-01

    This is an end-of-year report (Fiscal Year (FY) 2011) for the second year of effort on a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The second-year goals for this project included: (1) assessing the neutron source strength needed for the NRTA technique, (2) estimating count times, (3) assessing the effect of temperature on the transmitted signal, (4) estimating plutonium content in a spent fuel assembly, (5) providing a preliminary assessment of the neutron detectors, and (6) documenting this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes work performed over a nine month period from January-September 2011 and is to be considered a follow-on or add-on report to our previous published summary report from December 2010 (INL/EXT-10-20620).

  16. Spatial and energy distributions of skyshine neutron and gamma radiation from nuclear reactors on the ground-air boundary

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Y.; Netecha, M.E.; Vasiliev, A.P.; Avaev, V.N.; Vasiliev, G.A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Zelensky, D.I.; Istomin, Y.L.; Cherepnin, Y.S. [Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Semipalatinsk-21 (Kazakhstan); Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    A set of measurements on skyshine radiation was conducted at two special research reactors. A broad range of detectors was used in the measurements to record neutron and gamma radiations. Dosimetric and radiometric field measurements of the neutrons and gamma quanta of the radiation scattered in the air were performed at distances of 50 to 1000 m from the reactor during different weather conditions. The neutron spectra in the energy range of 1 eV to 10 MeV and the gamma quanta spectra in the range of 0.1-10 MeV were measured. (author)

  17. Evaluation of the total gamma-ray production cross-sections for nonelastic interaction of fast neutrons with iron nuclei

    International Nuclear Information System (INIS)

    Savin, M.V.; Nefedov, Yu.Ya; Livke, A.V.; Zvenigorodskij, A.G.

    2001-01-01

    Experimental data on the total gamma-ray production cross-sections for inelastic interaction of fast neutrons with iron nuclei were analysed. The total gamma-ray production cross-sections, grouped according to E γ , were evaluated in the neutron energy range 0.5-19 MeV. The statistical spline approximation method was used to evaluate the experimental data. Evaluated data stored in the ENDF, JENDL, BROND, and other libraries on gamma-ray production spectra and cross-sections for inelastic interaction of fast neutrons with iron nuclei, were analysed. (author)

  18. Applying thermal neutron radiography to non-destructive assays of dynamic systems

    International Nuclear Information System (INIS)

    Silvani, Maria I.; Almeida, Gevaldo L. de; Goncalves, Marcelo J.; Lopes, Ricardo T.

    2008-01-01

    Dynamic processes or systems frequently can not have their behavior directly analyzed due to safety reasons or because they require destructive assays, which can not be always afforded when high-cost equipment, devices and components are involved. Under these circumstances, some kind of non-destructive technique should be applied to preserve the safety of the personnel performing the assay, as well as the integrity of the piece being inspected. Thermal neutrons are specially suited as a tool for this purpose, thanks to their capability to pass through metallic materials, which could be utterly opaque to X-rays. This paper describes the accomplishments achieved at the Instituto de Engenharia Nuclear / CNEN, Brazil, aiming at the development of an Image Acquisition System capable to perform non-destructive assays using thermal neutrons. It is comprised of a thermal neutron source provided by the Argonauta research reactor, a converter-scintillating screen, and a CCD-based video camera optically coupled to the screen through a dark chamber equipped with a mirror. The developed system has been used to acquire 2D neutron radiographic images of static devices to reveal their inner structure, as well as movies of running systems and working devices to verify its functioning and soundness. Radiographic images of objects taken at different angles would be later on used as projections to retrieve - through a proper unfolding software - their 3D images expressed as attenuation coefficients for thermal neutrons. A quantitative performance of the system has been assessed through its Modulation Transfer Function - MTF. In order to determine this curve, unique collimators designed to simulate different spatial frequencies have been manufactured. Besides that, images of some objects have been acquired with the system being developed as well as using the conventional radiographic film, allowing thus a qualitative comparison between them. (author)

  19. Numerical Simulations of Pillar Structured Solid State Thermal Neutron Detector Efficiency and Gamma Discrimination

    Energy Technology Data Exchange (ETDEWEB)

    Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R

    2008-06-24

    This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.

  20. Alteration of UV primary fluorescence of vital tumor cells following irradiation with neutrons and gamma rays

    International Nuclear Information System (INIS)

    Merkle, K.

    1980-01-01

    The change of UV primary fluorescence intensity of vital unstained cells of Ehrlich ascites carcinoma after 60 Co-gamma and neutron irradiation was investigated. The mean neutron energy was 6.2 MeV. Fluorescence intensity was detected using impulse cytophotometry. The UV intensity of single cells was measured in the spectral range from 300-400 nm. An monotonous increase of dose-effect curves and a maximum at 3.5 Gy (neutrons) and 30 Gy (γ-rays) was obtained. The first relevant increase of fluorescence intensity was detected at 0.4 Gy (neutrons) and 0.75 Gy (γ-rays). Factors influencing the increase and decrease of primary fluorescence behavior of vital cells are discussed. (author)

  1. On the use of bismuth as a neutron and gamma ray filter

    International Nuclear Information System (INIS)

    Adib, M.; Kilany, M.

    2003-01-01

    A formula is given which, for neutron energies in the range 10 -4 < E<10 eV, permits calculation of the nuclear capture, thermal diffuse and bragg scattering cross-sections as a function of bismuth temperature crystalline form. Computer programs have been developed which allow calculations for the Bi rhombohedral structure in its poly-crystalline form and its equivalent hexagonal close-packed structure. Calculated total neutron cross-sections for poly-crystalline Bi at different temperatures were compared with measured values. Overall agreement is indicated between the formula fits and experimental data. Agreement was also obtained for values of Bi-single crystals, at room and liquid nitrogen temperatures. A feasibility study for use of Bi in powdered form, as a spread temperature and cutting plane for efficient transmission of thermal-reactor neutrons, and also for rejection of accompanying fast neutrons and gamma rays

  2. Digital pulse shape discrimination between fast neutrons and gamma rays with para-terphenyl scintillator

    Science.gov (United States)

    Chepurnov, A. S.; Kirsanov, M. A.; Klenin, A. A.; Klimanov, S. G.; Kubankin, A. S.

    2017-12-01

    In the presented work, we investigated several digital methods of a discrimination signals from fast neutrons and gamma quanta. The experimental setup consists of a Pu-Be neutron source, a scintillation detector with an organic para-terphenyl monocrystal, and a digitizer (CAEN DT5730, 500 MS/s). Mixed waveform sequences were stored and then separated by pulse shape. Four methods were used for signals separation. Comparison of the traditional and the new methods of Figure of Merit (FOM) calculation is given. FOM = 1.5 was obtained in our setup for the minimum threshold value. A scintillation detector with a para-terphenyl crystal was used to measure neutron yield in the neutron generator with carbon nanotubes.

  3. A low-background gamma-ray assay laboratory for activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Langland, J K [National Inst. of Standards and Technology, Gaithersburg, MD (USA). Center for Analytical Chemistry; Lindstrom, D J [National Aeronautics and Space Administration, Houston, TX (USA). Lyndon B. Johnson Space Center; Slaback, L A [National Inst. of Standards and Technology, Gaithersburg, MD (USA). Occupational Health and Safety Div.

    1990-12-20

    The sources of background in a gamma-ray detector were experimentally determined in underground and surface counting rooms, and an optimized shield was constructed at NIST. The optimum thickness of lead was 10-15 cm, with a greater thickness giving an increased background due to the buildup of tertiary cosmic-ray particles. Neither cadmium, tin, copper nor plastic (hydrocarbon or fluorocarbon) was desirable as a shield liner, since all these increased the background continuum or introduced characteristic peaks into the background spectrum. Two broad peaks in the background result from inelastic scattering of cosmic-ray neutrons (0.02 cm{sup -2} s{sup -1}) in germanium. These neutrons also excite the lower nuclear levels of lead and structural iron to produce additional gamma-ray peaks in the spectrum. The influence of the 20 MW NIST reactor, located 60 m from the detector, was undetectable. Comparisons among detectors and locations clearly separate cosmic from environmental components of the background. (orig.).

  4. Evaluation of the neutron self-interrogation approach for assay of plutonium in high materials

    International Nuclear Information System (INIS)

    Russo, P.A.; Menlove, H.O.; Fife, K.W.; West, M.H.

    1987-01-01

    The pyrochemical scrap recovery processes, designed to extract impurities from plutonium metal and compounds, generate a variety of plutonium-laden residues consisting of high (α,n) matrices of varying chemical composition, and often containing grams to tens of grams of americium. For such materials, multiplication corrections based on real neutron coincidence count rate, R, and total neutron count rate, T, measurements cannot be applied because of the large, unknown, and variable (α,n) component in the total neutron emission rate. A study of the prototype self-interrogation assay method is in progress at the Los Alamos plutonium facility. In the self-interrogation approach, the assay signature R(IF)/T is a function of effective fissile plutonium content, where R(IF) is the induced fission component of the measured reals rate, and T is the measured, (α,n)-dominated totals rate. The present study includes a calibration effort using standards consisting of mixtures of PuO 2 and PuF 4 in a salt-strip matrix. The neutron measurements of the standards and the process materials have been performed at the Los Alamos Plutonium Facility. The precision and accuracy of the self-interrogation method applied to pyrochemical residues is examined in this study

  5. Tidal heating and mass loss in neutron star binaries - Implications for gamma-ray burst models

    Science.gov (United States)

    Meszaros, P.; Rees, M. J.

    1992-01-01

    A neutron star in a close binary orbit around another neutron star (or stellar-mass black hole) spirals inward owing to gravitational radiation. We discuss the effects of tidal dissipation during this process. Tidal energy dissipated in the neutron star's core escapes mainly as neutrinos, but heating of the crust, and outward diffusion of photons, blows off the outer layers of the star. This photon-driven mass loss precedes the final coalescence. The presence of this eject material impedes the escape of gamma-rays created via neutrino interactions. If an e(+) - e(-) fireball, created in the late stages of coalescence, were loaded with (or surrounded by) material with the mean column density of the ejecta, it could not be an efficient source of gamma-rays. Models for cosmologically distant gamma-rays burst that involve neutron stars must therefore be anisotropic, so that the fireball expands preferentially in directions where the column density of previously blown-off material is far below the spherically averaged value which we have calculated. Some possible 'scenarios' along these lines are briefly discussed.

  6. Verification by the FISH translocation assay of historic doses to Mayak workers from external gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Sotnik, Natalia V.; Azizova, Tamara V. [Southern Urals Biophysics Institute (SUBI), Ozyorsk, Chelyabinsk Region (Russian Federation); Darroudi, Firouz [Leiden University Medical Center, Department of Toxicogenetics, Leiden (Netherlands); College of North Atlantic, Department of Health Science, Centre for Human Safety and Environmental Research, Doha (Qatar); Ainsbury, Elizabeth A.; Moquet, Jayne E.; Lloyd, David C.; Hone, Pat A.; Edwards, Alan A. [Public Health England, Chilton, Oxfordshire (United Kingdom); Fomina, Janna [Leiden University Medical Center, Department of Toxicogenetics, Leiden (Netherlands)

    2015-11-15

    The aim of this study was to apply the fluorescence in situ hybridization (FISH) translocation assay in combination with chromosome painting of peripheral blood lymphocytes for retrospective biological dosimetry of Mayak nuclear power plant workers exposed chronically to external gamma radiation. These data were compared with physical dose estimates based on monitoring with badge dosimeters throughout each person's working life. Chromosome translocation yields for 94 workers of the Mayak production association were measured in three laboratories: Southern Urals Biophysics Institute, Leiden University Medical Center and the former Health Protection Agency of the UK (hereinafter Public Health England). The results of the study demonstrated that the FISH-based translocation assay in workers with prolonged (chronic) occupational gamma-ray exposure was a reliable biological dosimeter even many years after radiation exposure. Cytogenetic estimates of red bone marrow doses from external gamma rays were reasonably consistent with dose measurements based on film badge readings successfully validated in dosimetry system ''Doses-2005'' by FISH, within the bounds of the associated uncertainties. (orig.)

  7. Measuring planetary neutron albedo fluxes by remote gamma-ray sensing

    International Nuclear Information System (INIS)

    Haines, E.L.; Metzger, A.E.

    1984-01-01

    A remote-sensing γ-ray spectrometer (GRS) is capable of measuring planetary surface composition through the detection of characteristic gamma rays. In addition, the planetary neutron leakage flux may be detected by means of a thin neutron absorber surrounding the γ-ray detector which converts the neutron flux into a γ-ray flux having a unique energy signature. The γ rays representing the neutron flux are observed against interference consisting of cosmic γ rays, planetary continuum and line emission, and a variety of gamma rays arising from cosmic-ray particle interactions with the γ-ray spectrometer and spacecraft (SC). In this paper the amplitudes of planetary and non-planetary neutron fluxes are assessed and their impact on the sensitivity of measurement is calculated for a lunar orbiter mission and a comet nucleus rendezvous mission. For a 100 h observation period from an altitude of 100 km, a GRS on a lunar orbiter can detect a thermal neutron albedo flux as low as 0.002 cm -2 s -1 and measure the expected flux of approx.=0.6 cm -2 s -1 with an uncertainty of 0.001 cm -2 s -1 . A GRS rendezvousing with a comet at a distance equal to the radius of the comet's nucleus, again for a 100 h observation time, should detect a thermal neutron albedo flux at a level of 0.006 cm -2 s -1 and measure the expected flux of approx.=0.4 cm -2 s -1 with an uncertainty of 0.004 cm -2 s -1 . Mapping the planetary neutron flux jointly with the direct detection of H will not only provide a more accurate model for translating observed γ-ray fluxes into concentrations but will also extend the effective sampling depth and should provide a capability for simple stratigraphic modeling of hydrogen. (orig.)

  8. {gamma} ray spectroscopy of neutron rich nuclei around N=20; Spectroscopie {gamma} des noyaux riches en neutrons autour de N=20

    Energy Technology Data Exchange (ETDEWEB)

    Gelin, M

    2007-09-15

    There is an island of inversion around {sup 32}Mg (12 protons, 20 neutrons) in contradiction with a shell closure N=20. It means a coexistence of spherical and deformed shapes. This work is devoted to the study of {gamma}-ray spectroscopy for nuclei in this region, based on an experiment done at GANIL with a composite secondary beam produced by fragmentation. The originality of the method used here lies in the possibility to study simultaneously several nuclei, and for each of them to explore several reaction channels. The VAMOS spectrometer was used for the identification of the ejectiles. The {gamma}-rays were detected with EXOGAM, a germanium clover array. The detectors used before and after the target allowed for a unique identification and a selection of the reaction channel: inelastic scattering, transfer and fragmentation reaction. In this thesis the following nuclei were studied: {sup 28}Ne, {sup 30-32}Mg {sup 31-34}Al, {sup 33-35}Si, {sup 35}P. New {gamma}-rays have been observed. The {gamma}-ray angular distributions and {gamma}-{gamma} angular correlations have been measured for some transitions. Assignment of spins and parities has been proposed for some states. In particular, in {sup 34}Si, the 3{sup -} assignment is confirmed and a new candidate for the second 0{sup +} has been proposed. In {sup 32}Mg, the state at 2.321 MeV, for which conflicting assignment existed, is deduced from the present data as a 4{sup +}, and a 6{sup +} state is proposed. (author)

  9. Inter-pulse high-resolution gamma-ray spectra using a 14 MeV pulsed neutron generator

    Science.gov (United States)

    Evans, L.G.; Trombka, J.I.; Jensen, D.H.; Stephenson, W.A.; Hoover, R.A.; Mikesell, J.L.; Tanner, A.B.; Senftle, F.E.

    1984-01-01

    A neutron generator pulsed at 100 s-1 was suspended in an artificial borehole containing a 7.7 metric ton mixture of sand, aragonite, magnetite, sulfur, and salt. Two Ge(HP) gamma-ray detectors were used: one in a borehole sonde, and one at the outside wall of the sample tank opposite the neutron generator target. Gamma-ray spectra were collected by the outside detector during each of 10 discrete time windows during the 10 ms period following the onset of gamma-ray build-up after each neutron burst. The sample was measured first when dry and then when saturated with water. In the dry sample, gamma rays due to inelastic neutron scattering, neutron capture, and decay were counted during the first (150 ??s) time window. Subsequently only capture and decay gamma rays were observed. In the wet sample, only neutron capture and decay gamma rays were observed. Neutron capture gamma rays dominated the spectrum during the period from 150 to 400 ??s after the neutron burst in both samples, but decreased with time much more rapidly in the wet sample. A signal-to-noise-ratio (S/N) analysis indicates that optimum conditions for neutron capture analysis occurred in the 350-800 ??s window. A poor S/N in the first 100-150 ??s is due to a large background continuum during the first time interval. Time gating can be used to enhance gamma-ray spectra, depending on the nuclides in the target material and the reactions needed to produce them, and should improve the sensitivity of in situ well logging. ?? 1984.

  10. Dosimetry of mixed gamma - neutron fluxes in the active zone of working reactor and gamma-flux after quenching

    International Nuclear Information System (INIS)

    Mussaeva, M.A.; Zinov'ev, V.; Ibragimova, E.M.; Muminov, M.I.

    2006-01-01

    Full text: For carrying out experiments in the channels of nuclear reactor, it is necessary to know the distribution of neutron flux and the intensity of accompanying gamma-radiation both in the working and quenched regimes. Dosimetric parameter of transparent dielectrics is based on the effect of monotonous changing of optical absorption or luminescence under neutrons and/or gamma-radiation. While the radioactivity induced in an element monitor is proportional only to a neutron fluence beginning from a threshold energy. Therefore the aim of this work was to determine the values of neutron and gamma-component fluxes separately and evaluate the contribution of each into the defect production in dielectrics. We used very pure quartz glass of KU-1 type, produced in Russian State Optical Institute by fusion from SiCl 4 in the mixed flow of O 2 +H 2 (impurities of Cl and OH up to 10 -2 % and the rest - below 10 -4 %), SiO 2 glasses with 30 % Ba, and also pure Ni wire. Since under irradiation in the working reactor samples were undergone mixed neutron and gamma fluxes, we suggested determination of intensity of gamma-radiation from radio-nuclides (products of uranium fission) after quenching the reactor by the current of ionization chamber and glass dosimeters. Samples of SiO 2 -BaO together with Ni monitors were irradiated for 1 hour in 18 channels of the active zone of the working reactor both in the sealed ampoules and in the contact with water of the 1-st cooling circuit at 40 deg C. The linear dependence of the induced optical density on the absorbed dose of n 0 + γ-radiation was obtained. Ni -monitors not sensitive to γ-radiation gained the induced radioactivity proportional to the absorbed energy of neutron flux above 1 MeV. Neutron fluxes in the 18 channels varied from 9.53·10 11 to 1.21·10 13 cm -2 s -1 corresponding to fluences from 3.43·10 15 to 4.3·10 16 cm -2 . Optical density of band 215 nm ascertained to E ' - center, which is ≡ Si * near oxygen

  11. Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gehrke, R.J.; Streier, G.G.

    1997-03-01

    During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from {approximately} 300 ppm to {approximately} 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of {approximately} 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants` Results; and Part 3, Evaluation of and Comments on Participants` Results.

  12. Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Streier, G.G.

    1997-03-01

    During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from ∼ 300 ppm to ∼ 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of ∼ 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants' Results; and Part 3, Evaluation of and Comments on Participants' Results

  13. Relative mass resolution technique for optimum design of a gamma nondestructive assay system

    International Nuclear Information System (INIS)

    Koh, Duck Joon

    1995-02-01

    Nondestructive assay(NDA) is a widely used nuclear technology for quantitative elemental and isotopic assay. Nondestructive assay is performed by the detection of an identifying radiation emerging from the sample, which can be unambiguously related to the element or isotope of interest. In every assay we can identify two distinct factors that lead to measurement uncertainty. We refer to these as statistical and spatial uncertainties. If the spatial distribution of the analyte and the matrix material in the sample are known and fairly constant from sample to sample, then the major source of measurement uncertainty is the statistical uncertainty resulting from randomness in the counting process. The spatial uncertainty is independent of the measurement time and therefore sets a lower limit to the measurement uncertainty, which is inherent in the assay system in conjunction with the population of samples to be measured. The only way to minimize the spatial uncertainty is an optimized design of the assay system. Therefore we have to decide on the type and number of detectors to be used, their deployment around the sample, the type of radiation to be measured, the duration of each measurement, the size and shape of the sample drum. The design procedure leading to the optimal assay system should be based on a quantitative(RMR:Relative Mass Resolution) comparison of the performance of each proposed design. For NDA system design of low level radwaste, a specific purpose Monte Carlo code has been developed to simulate point-source responses for sources within an assayed radwaste drum and to analyze the effect of scattered gammas from higher energy gammas on the spectrum of a low energy gamma-ray. We could use the well-known Monte Carlo code, such as MCNP for the simulation of NDA in the case of low level radwaste. But, MCNP is a multi-purpose Monte Carlo transport code for several geometries which requires large memory and long CPU time. For some cases in nuclear

  14. Feasibility Study On Using Crystalline Lead As a Neutron and Gamma Ray Filter

    International Nuclear Information System (INIS)

    Adib, M.; Naguib, K.; Ashry, A.; Fathalla, M.

    2000-01-01

    A generalized formula is given which allows to calculate the contribution of the total neutron cross- section including the Bragg scattering from different (hkI) planes to the neutron transmission through a solid crystalline material. The formula takes into account the crystalline form of the material (poly- or mono- crystal ) and crystal parameters. A computer program ISCANF-II was developed to provide the required calculations. The calculated values of the neutron transmission through a lead single crystal cut along the (311) plane were compared with the previously measured ones in the wavelength range 0.03-0.52 nm. The measured and calculated values were found to be in reasonable agreement within the statistical accuracy. The feasibility study on using a poly crystalline lead as a cold neutron filter and monocrystalline as a thermal neutron one is given. The optimum crystal thickness, temperature and characteristics for efficiently transmitting the thermal reactor neutrons, while removing simultaneously fast neutrons and gamma rays accompanying the thermal ones for the both cases are given

  15. Wholesomeness studies on gamma-irradiated smoked fish using short-term mutagenicity assays

    International Nuclear Information System (INIS)

    De la Rosa, A.M.; Banzon, R.B.

    1985-12-01

    The effect of gamma irradiation on the mutagenicity potential of wood-smoked mackerel (Rastrelliger sp.) was investigated. Smoked fish were irradiated with dose of 2.0, 4.0, 6.0 and 8.0 KGy, and tested for mutagenic activity using the Salmonella plate incorporation assay, host-mediated assay, and micronucleus test. The DMSO extract of unirradiated smoked fish was found to be mutagenic, without metabolic activation in Salmonella strains TA 100 and TA 104, both sensitive to base-pair substitution mutations. Strains TA 98 and TA 97 which are sensitive to frameshift mutations showed no mutagenic activity towards the same DMSO extract. The observed response towards the Salmonella strains was not affected by irradiation in the range of radiation doses studied. The presence of protamutagens in the DMSO extract of unirradiated smoked fish was not detected using the host-mediated assay. In another in-vivo test however, the same DMSO extract induced the formation of micronuclei in the bonemarrow cells of mice. Gamma irradiation up to a dose of 8.0 KGy did not affect the observed mutagenicity of wood-smoked fish. (author)

  16. Neutron and gamma probes: Their use in agronomy. Second edition

    International Nuclear Information System (INIS)

    2003-01-01

    Water is an essential requirement for life on the planet. It is often the single most limiting factor in crop and livestock production. Water is a scarce resource in many urban and rural environments worldwide. According to the FAO, the global demand for fresh water is doubling every 21 years. The quality of the finite water supplies is also under threat from industrial, agricultural and domestic sources of pollution. The majority of crops are grown under rain-fed conditions and adequate water supply is the main factor limiting crop production in semi-arid and sub-humid regions. On the other hand, currently 20% of the world's arable land is under irrigation providing 35 to 40% of all agricultural production. Irrigation mismanagement poses a serious threat to the environment through groundwater pollution and salinization. It is therefore, essential that water resources be used efficiently by regular monitoring of soil-water status in the unsaturated zone. The neutron depth probe, a nuclear-based technique, is utilized worldwide for this purpose. The neutron moisture gauge, since its introduction some 40 years ago, can now be considered a routine method in soil water studies. Many developments have since been introduced, in particular electronic components, which have significantly improved performance and expanded applications. Although the neutron scattering method is routinely utilised in many developed countries, its use is still limited in developing countries due to several factors. Neutron depth probes contain radioactive sources, which will present health and environmental hazards if a probe is improperly used, stored or disposed of. National and international legislation and regulations must be complied with. The strategic objective of the sub-program Soil and Water Management and Crop Nutrition of the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture is to develop and promote the adoption of nuclear-based technologies for optimising soil

  17. RPL-SC dosimetric system for measuring gamma and neutron irradiation in case of emergency

    International Nuclear Information System (INIS)

    Khristova, M. G.

    1993-01-01

    A RPL-SC dosimetric system is designed based on radiophotoluminescence (RPL) and on the effect of fast neutron bombardment of silicon semiconductor (SC) diodes. The experimental prototype consists of a computerized automatic measurement system and an individual dosimetric cassette accommodating RPL and SC detectors. The equipment includes: a device for measurement of the direct voltage of Si diodes and the RPL light emitted by RPL detectors; a compartment with dosimetric cassettes to be measured; a manipulator with three positions executing automatic measurement of cassettes; a computer and a printer. The system operates in both manual and automatic modes. In the manual mode each step of the manipulator is set up by the operator who changes the ranges after they have been filled to capacity and registers the results. In the automatic mode the whole process of maintaining the supply and control voltage, of manipulator's operation, measuring, data recording and data processing are controlled by a specially designed computer programme. Main technical parameters: 1) Measurement range of absorbed dose: gamma rays - 10 -3 to 10 2 Gy; thermal neutrons - 10 -3 to 10 2 Gy; fast neutrons - 10 to 30 Gy. 2) Energy range: gamma rays - 0.04 to 1.25 MeV; thermal neutrons - 0.024 eV; fast neutrons - 0.3 to 14 MeV. 3) Relative measurement error - ±15% 4) Recurrent measurement of one and the same dose. 5) Measurement time of 1 detector - 15 sec. (author)

  18. Fundamentals of passive nondestructive assay of fissionable material: laboratory workbook

    International Nuclear Information System (INIS)

    Reilly, T.D.; Augustson, R.H.; Parker, J.L.; Walton, R.B.; Atwell, T.L.; Umbarger, C.J.; Burns, C.E.

    1975-02-01

    This workbook is a supplement to LA-5651-M, ''Fundamentals of Passive Nondestructive Assay of Fissionable Material'' which is the text used during the Nondestructive Assay Training Session given by Group A-1 of the Los Alamos Scientific Laboratory. It contains the writeups used during the six laboratory sessions covering basic gamma-ray principles, quantitative gamma-ray measurements, uranium enrichment measurements, equipment holdup measurements, basic neutron principles, and quantitative neutron assay

  19. Fundamentals of passive nondestructive assay of fissionable material: laboratory workbook

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, T.D.; Augustson, R.H.; Parker, J.L. Walton, R.B.; Atwell, T.L.; Umbarger, C.J.; Burns, C.E.

    1975-02-01

    This workbook is a supplement to LA-5651-M, ''Fundamentals of Passive Nondestructive Assay of Fissionable Material'' which is the text used during the Nondestructive Assay Training Session given by Group A-1 of the Los Alamos Scientific Laboratory. It contains the writeups used during the six laboratory sessions covering basic gamma-ray principles, quantitative gamma-ray measurements, uranium enrichment measurements, equipment holdup measurements, basic neutron principles, and quantitative neutron assay.

  20. Assessment of gamma ray-induced DNA damage in Lasioderma serricorne using the comet assay

    International Nuclear Information System (INIS)

    Kameya, Hiromi; Miyanoshita, Akihiro; Imamura, Taro; Todoriki, Setsuko

    2012-01-01

    We attempted a DNA comet assay under alkaline conditions to verify the irradiation treatment of pests. Lasioderma serricorne (Fabricius) were chosen as test insects and irradiated with gamma rays from a 60 Co source at 1 kGy. We conducted the comet assay immediately after irradiation and over time for 7 day. Severe DNA fragmentation in L. serricorne cells was observed just after irradiation and the damage was repaired during the post-irradiation period in a time-dependent manner. The parameters of the comet image analysis were calculated, and the degree of DNA damage and repair were evaluated. Values for the Ratio (a percentage determined by fluorescence in the damaged area to overall luminance, including intact DNA and the damaged area of a comet image) of individual cells showed that no cells in the irradiated group were included in the Ratio<0.1 category, the lowest grade. This finding was observed consistently throughout the 7-day post-irradiation period. We suggest that the Ratio values of individual cells can be used as an index of irradiation history and conclude that the DNA comet assay under alkaline conditions, combined with comet image analysis, can be used to identify irradiation history. - Highlights: ► We investigated the DNA comet assay to verify the irradiation of pests. ► Ratio and Tail Moment were higher in irradiated groups than in the control group. ► The DNA comet assay can be used to identify irradiation history.

  1. Application of neutron/gamma transport codes for the design of explosive detection systems

    International Nuclear Information System (INIS)

    Elias, E.; Shayer, Z.

    1994-01-01

    Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs

  2. Magnetization of neutron star matter and implications in physics of soft gamma repeaters

    Energy Technology Data Exchange (ETDEWEB)

    Kondratyev, V N [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-01-01

    The magnetization of neutron star matter is considered within the thermodynamic formalism. The quantization effects are demonstrated to result in sharp abrupt magnetic field dependence of nuclide magnetic moments. Accounting for inter-nuclide magnetic coupling we show that such anomalies give rise to erratic jumps in magnetotransport of neutron star crusts. The properties of such a noise are favorably compared with burst statistics of Soft Gamma Repeaters. PACS: 97.60.Jd, 21.10.Dr, 26.60.+c, 95.30.Ky. (author)

  3. Measurement of log moisture content and density by gamma and neutron backscatter

    International Nuclear Information System (INIS)

    Barry, B.J.

    2002-01-01

    Measurement of the moisture content and green density of wood was investigated using scattering of gamma rays and neutrons. Both of these processes are dependent on density but neutrons are particularly sensitive to the hydrogen content, which changes with moisture content. A material mimicking the green density and moisture content of real wood was successfully used in a laboratory study to establish the feasibility of measuring these within the range found in standing trees. A later field trial indicated that the technique needed more development to take account of the natural variability of real trees. (author). 3 refs., 11 figs., 1 table

  4. A new digital method for high precision neutron-gamma discrimination with liquid scintillation detectors

    International Nuclear Information System (INIS)

    Nakhostin, M

    2013-01-01

    A new pulse-shape discrimination algorithm for neutron and gamma (n/γ) discrimination with liquid scintillation detectors has been developed, leading to a considerable improvement of n/γ separation quality. The method is based on triangular pulse shaping which offers a high sensitivity to the shape of input pulses, as well as, excellent noise filtering characteristics. A clear separation of neutrons and γ-rays down to a scintillation light yield of about 65 keVee (electron equivalent energy) with a dynamic range of 45:1 was achieved. The method can potentially operate at high counting rates and is well suited for real-time measurements.

  5. Neutron-gamma discrimination employing pattern recognition of the signal from liquid scintillator

    CERN Document Server

    Kamada, K; Ogawa, S

    1999-01-01

    A pattern recognition method was applied to the neutron-gamma discrimination of the pulses from the liquid scintillator, NE-213. The circuit for the discrimination is composed of A/D converter, fast SCA, memory control circuit, two digital delay lines and two buffer memories. All components are packed on a small circuit board and are installed into a personal computer. Experiments using a weak sup 2 sup 5 sup 2 Cf n-gamma source were undertaken to test the feasibility of the circuit. The circuit is of very easy adjustment and, at the same time, of very economical price when compared with usual discrimination circuits, such as the TAC system.

  6. Neutron/gamma-ray techniques for investigating the deterioration of historic buildings

    International Nuclear Information System (INIS)

    Evans, L.G.; Trombka, J.I.

    1986-01-01

    The degradation of building materials is a major problem for the preservation of historic structures. The presence of contaminants in the constituent materials is often a cause of the deterioration. Neutron-induced, prompt gamma-ray techniques for nondestructive elemental analysis are used to determine the distribution of contaminants in building walls. The application of these methods for the diagnosis of an 18th century historic building indicates that the distributions within the building walls of moisture, salt and bulk density can be obtained. The results of an analysis of the gamma-ray spectra are confirmed by independent measurements on two sample cores taken through one wall. (orig.)

  7. Method of assaying uranium with prompt fission and thermal neutron borehole logging adjusted by borehole physical characteristics

    International Nuclear Information System (INIS)

    Barnard, R.W.; Jensen, D.H.

    1982-01-01

    Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or eqithermal dieaway. Various calibration factors enhance the accuracy of the measurement

  8. Monte Carlo neutron and gamma-ray calculations

    International Nuclear Information System (INIS)

    Mendelsohn, Edgar

    1987-01-01

    Kerma in tissue and the activation produced in sulfur and cobalt due to prompt neutrons from the Hiroshima and Nagasaki bombs were calculated out to 2000 m from the hypocenter in 100 m increments. As neutron sources weapon output spectra calculated by investigators from the Los Alamos National Laboratory (LANL) were used. Other parameters, such as burst height and air and ground densities and compositions, were obtained from recent sources. The LLNL Monte Carlo transport code TART was used for these calculations. TART accesses the well-established 1985 ENDL cross-section library, which has built-in reaction cross sections. The zoning for this problem was a full two-dimensional geometry with a ceiling height of 1100 m and a ground thickness of 30 cm. For the Hiroshima calculations (including sulfur activation) and untilted source was used. However, a special sulfur activation problem using a source tilted 15 deg was run for which the ratios to the untilted case are reported. The TART code uses a technique for solving the transport equation that is different from that of the ORNL DOT code; it also draws on a specially evaluated cross-section library (ENDL) and uses a larger group structure than DOT. One of the purposes of this work was to instill confidence in the DOT calculations that will be used directly in the dose reassessment of A-bomb survivors. The TART results were compared with values calculated with the DOT code by investigators from ORNL and found to be in good agreement for the most part. However, the sulfur activation comparison is disappointing. Because the sulfur activation is caused by higher energy neutrons (which should have experienced fewer collisions than those causing cobalt activation, for example), better agreement than what is reported here would be expected

  9. Non-invasive analysis of industrial products using the simultaneous transmission of neutrons and gamma rays (Neugat) method

    International Nuclear Information System (INIS)

    Bartle, C.M.

    1998-01-01

    This research programme is designed to develop industrial measurement systems utilising simultaneous transmission of neutrons and gamma rays (Neugat method). Descriptions of these systems have been given in reports and magazine articles, and industrial site trials have been undertaken. (author)

  10. Attenuation of Neutron and Gamma Radiation by a Composite Material Based on Modified Titanium Hydride with a Varied Boron Content

    Science.gov (United States)

    Yastrebinskii, R. N.

    2018-04-01

    The investigations on estimating the attenuation of capture gamma radiation by a composite neutron-shielding material based on modified titanium hydride and Portland cement with a varied amount of boron carbide are performed. The results of calculations demonstrate that an introduction of boron into this material enables significantly decreasing the thermal neutron flux density and hence the levels of capture gamma radiation. In particular, after introducing 1- 5 wt.% boron carbide into the material, the thermal neutron flux density on a 10 cm-thick layer is reduced by 11 to 176 factors, and the capture gamma dose rate - from 4 to 9 times, respectively. The difference in the degree of reduction in these functionals is attributed to the presence of capture gamma radiation in the epithermal region of the neutron spectrum.

  11. Focussing X-rays, gamma rays and neutrons

    International Nuclear Information System (INIS)

    Smither, R.K.

    1982-01-01

    A diffraction crystal or grating has a face for receiving a beam of photons or neutrons and diffraction planar spacing or grating spacing along that face with the spacing increasing progressively along the face to provide a decreasing Bragg diffraction angle for a monochromatic radiation and thereby increasing the usable area and acceptance angle. The increased planar spacing for a diffraction crystal is provided by the use of a temperature differential across the crystalline structure, by assembling a plurality of crystalline structures with different compositions, by an individual crystalline structure with a varying composition and thereby a changing planar spacing along its face, and by combinations of these techniques. (author)

  12. The generation, validation and testing of a coupled 219-group neutron 36-group gamma ray AMPX-II library

    International Nuclear Information System (INIS)

    Panini, G.C.; Siciliano, F.; Lioi, A.

    1987-01-01

    The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations

  13. A programme for Euratom safeguards inspectors, used in the assay of plutonium bearing materials by passive neutron interrogation

    International Nuclear Information System (INIS)

    Vocino, V.; Farese, N.; Maucq, T.; Nebuloni, M.

    1991-01-01

    The programme PECC (Passive Euratom Coincidence Counters) has been developed at the Joint Research Center, Ispra by the Euratom Safeguards Directorate, Luxembourg and the Safety Technology Institute, Ispra for the acquisition, evaluation, management and storage of measurements data originating from passive neutron assay of plutonium bearing materials. The software accommodates the implementation of the NDA (Non Destructive Assay) procedures for all types of passive neutron coincidence deployed by the Euratom Safeguards Directorate, Luxembourg

  14. The Monte Carlo simulation of the neutron-induced prompt gamma ray spectroscopy of the CW abandoned by Japan

    International Nuclear Information System (INIS)

    Wang Bairong; Yang Zhongping; Zhan Wenzhong

    2003-01-01

    This paper introduced the principle of identifying the chemical weapon abandoned by Japan by neutron-induced prompt gamma ray. Using the MCNP-4C Monte Carlo program, this paper simulated and analyzed the neutron-induced prompt gamma ray spectroscopy of chemical weapon abandoned by Japan, whereby supply important datum and reference for the aftertime deeper research and disposal of Japan-abandoned chemical weapon. (authors)

  15. Gamma-ray production cross sections for 0.9 to 20 MeV neutron interactions with 10B

    International Nuclear Information System (INIS)

    Bywater, R.L. Jr.

    1986-09-01

    Gamma-ray spectral data previously obtained at the 20-meter station of the Oak Ridge Electron Linear Accelerator flight-path 8 were studied to determine cross sections for 0.9- to 20-MeV neutron interactions with 10 B. Data reduction techniques, including those for determination of incident neutron fluences as well as those to compensate for Doppler-broadened gamma-ray-detection responses, are given in some detail in this report. 9 refs., 4 figs., 2 tabs

  16. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lamadrid Boada, Ana I; Garcia Lima, Omar [Centro de Proteccion e Higiene de las Radiaciones (CPHR), La Habana (Cuba); Delbos, Martine; Voisin, Philipe; Roy, Laurence [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)

    2008-07-01

    Dose-effect curves for dose assessment in Gamma and neutron overexposures to high doses are presented in this paper for the first time in literature. The relationships were obtained by plotting the Premature Chromosome Condensation -rings (PCC-R) frequencies in PCC lymphocytes obtained by chemical induction with Calyculin A in vitro, with radiation doses between 5 to 25 Gy. For the elaboration of these curves 9 676 PCC cells in G1 G2 and M stages were analyzed. The results were fitted to a lineal quadratic model in Gamma irradiation and showed saturation starting from 20 Gy. For neutron irradiation the data was fitted to a lineal quadratic model up to 10 Gy, and then a markedly cell cycle arrest and saturation was observed. These curves are of particular interest for victims exposed to doses exceeding 5 Gy where it is always very difficult to estimate a dose using the conventional technique. (author)

  17. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lamadrid B, A I; Garcia L, O [CPHR, Calle 20 No. 4113 e/41 y 47, Playa, La Habana 11300 (Cuba); Delbos, M; Voisin, P; Roy, L [Institut de Radioprotection et de Surete Nucleaire, BP 17, 92262 Fontenay-aux-Roses (France)

    2006-07-01

    Dose-effect curves for dose assessment in Gamma and neutron overexposures to high doses are presented in this paper for the first time in literature. The relationships were obtained by plotting the Premature Chromosome Condensation -rings (PCC-R) frequencies in PCC Iymphocytes obtained by chemical induction with Calyculin A in vitro, with radiation doses between 5 to 25 Gy. For the elaboration of these curves 9 676 PCC cells in Gl G2 and M stages were analyzed. The results were fitted to a lineal quadratic model in Gamma irradiation. For neutron irradiation the data was fitted to a lineal quadratic model up to 10 Gy and then a markedly cell cycle arrest and saturation was observed. These curves are of particular interest for victims exposed to doses exceeding 5 Gy where it is always very difficult to estimate a dose using the conventional technique. (Author)

  18. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation

    International Nuclear Information System (INIS)

    Lamadrid B, A.I.; Garcia L, O.; Delbos, M.; Voisin, P.; Roy, L.

    2006-01-01

    Dose-effect curves for dose assessment in Gamma and neutron overexposures to high doses are presented in this paper for the first time in literature. The relationships were obtained by plotting the Premature Chromosome Condensation -rings (PCC-R) frequencies in PCC Iymphocytes obtained by chemical induction with Calyculin A in vitro, with radiation doses between 5 to 25 Gy. For the elaboration of these curves 9 676 PCC cells in Gl G2 and M stages were analyzed. The results were fitted to a lineal quadratic model in Gamma irradiation. For neutron irradiation the data was fitted to a lineal quadratic model up to 10 Gy and then a markedly cell cycle arrest and saturation was observed. These curves are of particular interest for victims exposed to doses exceeding 5 Gy where it is always very difficult to estimate a dose using the conventional technique. (Author)

  19. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation

    International Nuclear Information System (INIS)

    Lamadrid Boada, Ana I.; Garcia Lima, Omar; Delbos, Martine; Voisin, Philipe; Roy, Laurence

    2008-01-01

    Dose-effect curves for dose assessment in Gamma and neutron overexposures to high doses are presented in this paper for the first time in literature. The relationships were obtained by plotting the Premature Chromosome Condensation -rings (PCC-R) frequencies in PCC lymphocytes obtained by chemical induction with Calyculin A in vitro, with radiation doses between 5 to 25 Gy. For the elaboration of these curves 9 676 PCC cells in G1 G2 and M stages were analyzed. The results were fitted to a lineal quadratic model in Gamma irradiation and showed saturation starting from 20 Gy. For neutron irradiation the data was fitted to a lineal quadratic model up to 10 Gy, and then a markedly cell cycle arrest and saturation was observed. These curves are of particular interest for victims exposed to doses exceeding 5 Gy where it is always very difficult to estimate a dose using the conventional technique. (author)

  20. American National Standard: neutron and gamma-ray flux-to-dose rate factors

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This Standard presents data recommended for computing biological dose rates due to neutron and gamma-ray radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are given; the energy range for the gamma-ray conversion factors is 0.01 to 15 MeV. Specifically, this Standard is intended for use by shield designers to calculate wholebody dose rates to radiation workers and the general public. Establishing dose-rate limits is outside the scope of this Standard. Use of this Standard in cases where the dose equivalents are far in excess of occupational exposure guidelines is not recommended

  1. Simulation of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation

    International Nuclear Information System (INIS)

    Kluson, J.; Jansky, B.

    2009-01-01

    Reference mixed neutron-gamma fields are used for test and calibration of dosimetric and spectrometric systems, intercomparison measurements, and benchmark tests and represent experimental base for reactor studies. Set of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation was build in the NRI Rez. Extended sets of measurements and simulation calculations were done to describe the reference mixed field dosimetry and spectral characteristics with best achievable precision. The Monte Carlo technique was used for different experimental setups models description, comparison and verification and field characteristics simulation. Effects (hardly distinguishable experimentally) were also studied ( contributions from individual parts of experimental setup, field individual components and next effects as shadow shield cones transparency, etc.). Some results and main conclusions of these studies and calculations are presented and discussed. (authors)

  2. Isozymes variability in rice mutants induced by fast neutrons and gamma rays

    International Nuclear Information System (INIS)

    Fuentes, J.L.; Alvarez, A.; Gutierrez, L.; Deus, J.E.

    2001-01-01

    The isozyme variability of a group of rice mutants induced through gamma and fast neutron (14 MeV) irradiation was studied. Polymorphisms were detected using esterase, peroxidase, polyphenol oxidase and alcohol dehydrogenase systems. The mean value of genetic similarity among the different cultivars, which arose from isozymes, was 0.75. The dendrogram was constructed based on genetic similarity matrices, designed with isozyme data using the unweighed pair group method arithmetic average (UPGMA) method. The efficiency of the UPGMA model for the estimation of genetic relationship among cultivars was supported by cophenetic correlation coefficients. Such values indicate that the distortion degree for the estimated similarities was minimal. It was found that both gamma rays and fast neutrons generated a wide range of variability which can be detected by means of isozyme patterns, even in closely related cultivars. (author)

  3. Isozymes variability in rice mutants induced by fast neutrons and gamma rays

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes, J L; Alvarez, A [Centro de Estudios Aplicados al Desarrollo Nuclear (CEADEN), Miramar, Playa, Havana (Cuba); Gutierrez, L; Deus, J E [Instituto de Investigaciones del Arroz, Bauta, Havana (Cuba)

    2001-05-01

    The isozyme variability of a group of rice mutants induced through gamma and fast neutron (14 MeV) irradiation was studied. Polymorphisms were detected using esterase, peroxidase, polyphenol oxidase and alcohol dehydrogenase systems. The mean value of genetic similarity among the different cultivars, which arose from isozymes, was 0.75. The dendrogram was constructed based on genetic similarity matrices, designed with isozyme data using the unweighed pair group method arithmetic average (UPGMA) method. The efficiency of the UPGMA model for the estimation of genetic relationship among cultivars was supported by cophenetic correlation coefficients. Such values indicate that the distortion degree for the estimated similarities was minimal. It was found that both gamma rays and fast neutrons generated a wide range of variability which can be detected by means of isozyme patterns, even in closely related cultivars. (author)

  4. Simulation of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation

    International Nuclear Information System (INIS)

    Kluson, J.; Jansky, B.

    2008-01-01

    Reference mixed neutron-gamma fields are used for test and calibration of dosimetric and spectrometric systems, intercomparison measurements, and benchmark tests and represent experimental base for reactor studies. Set of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation was build in the NRI Rez. Extended sets of measurements and simulation calculations were done to describe the reference mixed field dosimetry and spectral characteristics with best achievable precision. The Monte Carlo technique was used for different experimental setups models description, comparison and verification and field characteristics simulation. Effects (hardly distinguishable experimentally) were also studied ( contributions from individual parts of experimental setup, field individual components and next effects as shadow shield cones transparency, etc.). Some results and main conclusions of these studies and calculations are presented and discussed. (authors)

  5. MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications

    International Nuclear Information System (INIS)

    Selcow, E.C.; McKinney, G.W.

    2000-01-01

    The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90

  6. Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.

    Science.gov (United States)

    Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

    2017-06-01

    The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. The effect of gamma and fast neutron irradiations on M1 seedling growth in soybean

    International Nuclear Information System (INIS)

    Hassan, S.; Mohammad, T.; Khan, S.

    1985-01-01

    Seeds of three varieties of soybean, i.e. Bragg, Hodgson and Lee-74, having a moisture content of 11-13% were irradiated with doses of gamma, 100,200,300,400 and 500 Gray and fast neutron, 5,10,20,25 and 30 Gray, to study the effect on M1 seedling growth. The parameters studied were germination, seedling height and epicotyl length. Growth inhibition was found to increase with increasing radiation doses and the effect on germination was observed only at higher doses. Among early assessable M1 parameters for radio-sensitivity, epicotyl length has proved to be most sensitive, and hence most useful. The Relative Biological Effectiveness (RBE) values for the three varieties differed slightly for epicotyl length and the difference was more pronounced for seedling height. A dose range of 150-300 Gray of gamma rays and 10-15 Gray of fast neutron might prove useful for efficient induced mutation. (authors)

  8. Accuracy and borehole influences in pulsed neutron gamma density logging while drilling

    Energy Technology Data Exchange (ETDEWEB)

    Yu Huawei [College of Geo-Resources and Information, China University of Petroleum, Qingdao, Shandong 266555 (China); Center for Engineering Applications of Radioisotopes (CEAR), Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Sun Jianmeng [College of Geo-Resources and Information, China University of Petroleum, Qingdao, Shandong 266555 (China); Wang Jiaxin [Center for Engineering Applications of Radioisotopes (CEAR), Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Gardner, Robin P., E-mail: gardner@ncsu.edu [Center for Engineering Applications of Radioisotopes (CEAR), Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2011-09-15

    A new pulsed neutron gamma density (NGD) logging has been developed to replace radioactive chemical sources in oil logging tools. The present paper describes studies of near and far density measurement accuracy of NGD logging at two spacings and the borehole influences using Monte-Carlo simulation. The results show that the accuracy of near density is not as good as far density. It is difficult to correct this for borehole effects by using conventional methods because both near and far density measurement is significantly sensitive to standoffs and mud properties. - Highlights: > Monte Carlo evaluation of pulsed neutron gamma-ray density tools. > Results indicate sensitivity of the tool to standoff and mudcake properties. > Accuracy of far spaced detector is better than near spaced.

  9. Accuracy and borehole influences in pulsed neutron gamma density logging while drilling

    International Nuclear Information System (INIS)

    Yu Huawei; Sun Jianmeng; Wang Jiaxin; Gardner, Robin P.

    2011-01-01

    A new pulsed neutron gamma density (NGD) logging has been developed to replace radioactive chemical sources in oil logging tools. The present paper describes studies of near and far density measurement accuracy of NGD logging at two spacings and the borehole influences using Monte-Carlo simulation. The results show that the accuracy of near density is not as good as far density. It is difficult to correct this for borehole effects by using conventional methods because both near and far density measurement is significantly sensitive to standoffs and mud properties. - Highlights: → Monte Carlo evaluation of pulsed neutron gamma-ray density tools. → Results indicate sensitivity of the tool to standoff and mudcake properties. → Accuracy of far spaced detector is better than near spaced.

  10. A fast Monte Carlo program for pulsed-neutron capture-gamma tools

    International Nuclear Information System (INIS)

    Hovgaard, J.

    1992-02-01

    A fast model for the pulsed-neutron capture-gamma tool has been developed. It is believed that the program produce valid results even though some approximation have been introduced. A correct γ photon transport simulation, which is under preparation, has for instance not yet been included. Simulations performed so far has shown that the model, with respect to computing time and accuracy, fully lives up to expectations with respect to computing time and accuracy. (au)

  11. Systematic Assessment of Neutron and Gamma Backgrounds Relevant to Operational Modeling and Detection Technology Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hornback, Donald Eric [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Johnson, Jeffrey O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nicholson, Andrew D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peplow, Douglas E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ayaz-Maierhafer, Birsen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This report summarizes the findings of a two year effort to systematically assess neutron and gamma backgrounds relevant to operational modeling and detection technology implementation. The first year effort focused on reviewing the origins of background sources and their impact on measured rates in operational scenarios of interest. The second year has focused on the assessment of detector and algorithm performance as they pertain to operational requirements against the various background sources and background levels.

  12. Tests on a digital neutron-gamma pulse shape discriminator with NE213

    International Nuclear Information System (INIS)

    Bell, Z.W.

    1981-01-01

    A technique using charge sensitive analog-to-digital converters to do neutron-gamma pulse shape discrimination is reported. The converters are gated by short (135 ns) pulses so as to reduce pile-up and the timing is such that the slow and total light output from the scintillator are measured. Preliminary tests indicate that the system performs reasonably well but poorer than some reported analog systems employing gated integrators or cross-over techniques. (orig.)

  13. Neutron and gamma-ray sources in LWR high-level nuclear waste

    International Nuclear Information System (INIS)

    Dupree, S.A.

    1977-06-01

    Predictions of the composition of high-level waste from U-fueled LWRs have been used to calculate the neutron and gamma-ray sources in such waste at cooling times of 3 and 10 years. The results are intended for interim application to studies of waste shipping and storage pending the availability of more exact knowledge of fuel recycling and of waste concentration and solidification

  14. Quantitative and qualitative applications of the neutron-gamma borehole logging

    International Nuclear Information System (INIS)

    Charbucinski, J.; Eisler, P.L.; Borsaru, M.; Aylmer, J.A.

    1990-01-01

    Two examples of neutron-gamma borehole logging application are described. In the quantitative application of the PGNAA technique, research was carried out both in the laboratory and at a mine to establish a suitable borehole logging technology for Mn-grade predictions. As an example of qualitative application of PGNAA, use of this method has been demonstrated for determination of lithology. (author). 4 refs, 10 figs, 7 tabs

  15. Design of an automatic sample changer for the measurement of neutron flux by gamma spectrometry

    International Nuclear Information System (INIS)

    Gago, Javier; Bruna, Ruben; Baltuano, Oscar; Montoya, Eduardo; Descreaux, Killian

    2014-01-01

    This paper presents calculus, selection and components design for the construction of an automatic system in order to measure neutron flux in a working nuclear reactor by the gamma spectrometry technique using samples irradiated on the RP-10 nucleus. This system will perform the measurement of interchanging 100 samples in a programed and automatic way, reducing operation time by the user and obtaining more accurate measures. (authors).

  16. Application of the neutron gamma method to a study of water seepage under a rice plantation

    International Nuclear Information System (INIS)

    Puard, M.; Couchat, P.; Moutonnet, P.

    1980-01-01

    In order to determine the share of percolation in the pollution by pesticides (particularly Lindane) being carried down in the drainage water of rice plantations, an application of the neutron gamma method under rice cultivation in the Camargue is suggested. A preliminary laboratory study enabled a comparison to be made between deuteriated water (DHO) and tritiated water (THO) used as water tracers in the determination of the dispersive phenomena and retention in a column of saturated soil [fr

  17. Isotopic composition of uranium in U3O8 by neutron induced reactions utilizing thermal neutrons from critical facility and high resolution gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Acharya, R.; Pujari, P.K.; Goel, Lokesh

    2015-01-01

    Uranium in oxide and metal forms is used as fuel material in nuclear power reactors. For chemical quality control, it is necessary to know the isotopic composition (IC) of uranium i.e., 235 U to 238 U atom ratio as well as 235 U atom % in addition to its total concentration. Uranium samples can be directly assayed by passive gamma ray spectrometry for obtaining IC by utilizing 185 keV (γ-ray abundance 57.2%) of 235 U and 1001 keV (γ-ray abundance 0.837%) of 234m Pa (decay product of 238 U). However, due to low abundance of 1001 keV, often it is not practiced to obtain IC by this method as it gives higher uncertainty even if higher mass of sample and counting time are used. IC of uranium can be determined using activity ratio of neutron induced fission product of 235 U to activation product of 238 U ( 239 Np). In the present work, authors have demonstrated methodologies for determination of IC of U as well as 235 U atom% in natural ( 235 U 0.715%) and low enriched uranium (LEU, 3-20 atom % of 235 U) samples of uranium oxide (U 3 O 8 ) by utilizing ratio of counts at 185 keV γ-ray or γ-rays of fission products with respect to 277 keV of 239 Np. Natural and enriched samples (about 25 mg) were neutron irradiated for 4 hours in graphite reflector position of AHWR Critical Facility (CF) using highly thermalized (>99.9% thermal component) neutron flux (∼10 7 cm -2 s -1 )

  18. Comparison of results of assaying and neutron activation analysis when determining gold and silver content

    International Nuclear Information System (INIS)

    Vaganov, P.A.; Bulnaev, A.I.; Kulikov, V.D.; Mejer, V.A.; Zakharevich, K.V.

    1977-01-01

    Compared are results of simultaneous determination of gold and silver content in rock samples by the methods of neutron activation analysis and assaying. Rock samples were irradiated by thermal neutron flux of 5x10 13 nxcm -2 xs -1 during 12 hours. The gold content was determined in 8-12 days after irradiation, and silver content in 40-50 days. T he gold content determination was performed by 411.8 keV γ quanta of 198 Au. To establish the silver content two analytical lines of sup(110m)Ag isomer with the energy of 657.7 and 937.4 keV were used. The sensitivity threshold of Au content determination amounts to 3x10 -6 % (or 1x10 -9 g) and that for Ag is 2x10 -40 % (using γ line with the energy of 657.7 keV). The comparison of the results of assaying and neutron-activation analysis has shown for silver a good agreement between the both methods, the coefficient of pair correlation being equal to 0.997. For gold the divergence between the methods is observed. The activation analysis provides on the average lower values of gold content

  19. Portable gamma and thermal neutron probe using a 6LiI(Eu) crystal

    International Nuclear Information System (INIS)

    Carneiro, Clemente J.G.; Araujo, Geraldo P.; Milian, Felix M.; Barbosa, Jurandir C.; Garcia, Fermin; Oliveira, Arno H.; Silva, Mario R.S.; Penna, Rodrigo

    2011-01-01

    Europium-activated lithium-6 iodide is a scintillator used for gamma and neutron counting. A portable detection system was built based on this scintillator. This system has three modules: the scintillator, a 10 m liquid light guide, and a Hamamatsu photon counting head H9319 used as a light pulse digitizer. Data transfer, measurement time and other necessary adjustment can be controlled by software from the PC through the RS-232C interface. The scintillator, a crystal of 6 LiI(Eu), is a small cylinder with 3 mm diameter and 40 mm length completely sealed in an aluminum tube coupled to the light guide. The small size of the scintillator increases the neutron/gamma count ratio, since 2 to 3 mm of thickness of this crystal absorbs all thermal neutrons. Intensities of X and gamma rays, and thermal neutrons can be recorded for time intervals of 10 ms to 1 s storing up to 10000 countings. The system was calibrated for measuring radiation doses for validating numerical models in dosimetry. Two characteristic reinforce this application, measurements can be done at several meters away from the radiation source and also inside of water. In addition, it was used to build nuclear probes based on Compton scattering or neutron moderation in porous media by attaching an AmBe source to the top of the aluminum tube. Tests were done to determine the reproducibility of counting rates. Background counting was measured at several temperatures to verify the influence of dark current of PMT. Sealed AmBe, low activity Am, and X rays sources were used for studies of radiation counting statistics. X rays apparatus was used to correlate counting rates measured with the 6 LiI(Eu) detection system and doses measured with an ionization chamber at several distances from the X ray source. (author)

  20. Portable gamma and thermal neutron probe using a {sup 6}LiI(Eu) crystal

    Energy Technology Data Exchange (ETDEWEB)

    Carneiro, Clemente J.G.; Araujo, Geraldo P.; Milian, Felix M.; Barbosa, Jurandir C.; Garcia, Fermin [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Centro de Pesquisas em Ciencias e Tecnologias das Radiacoes (CPqCTR); Oliveira, Arno H.; Silva, Mario R.S.; Penna, Rodrigo [Universidade Federal de Minas Gerais (DEN-UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    Europium-activated lithium-6 iodide is a scintillator used for gamma and neutron counting. A portable detection system was built based on this scintillator. This system has three modules: the scintillator, a 10 m liquid light guide, and a Hamamatsu photon counting head H9319 used as a light pulse digitizer. Data transfer, measurement time and other necessary adjustment can be controlled by software from the PC through the RS-232C interface. The scintillator, a crystal of {sup 6}LiI(Eu), is a small cylinder with 3 mm diameter and 40 mm length completely sealed in an aluminum tube coupled to the light guide. The small size of the scintillator increases the neutron/gamma count ratio, since 2 to 3 mm of thickness of this crystal absorbs all thermal neutrons. Intensities of X and gamma rays, and thermal neutrons can be recorded for time intervals of 10 ms to 1 s storing up to 10000 countings. The system was calibrated for measuring radiation doses for validating numerical models in dosimetry. Two characteristic reinforce this application, measurements can be done at several meters away from the radiation source and also inside of water. In addition, it was used to build nuclear probes based on Compton scattering or neutron moderation in porous media by attaching an AmBe source to the top of the aluminum tube. Tests were done to determine the reproducibility of counting rates. Background counting was measured at several temperatures to verify the influence of dark current of PMT. Sealed AmBe, low activity Am, and X rays sources were used for studies of radiation counting statistics. X rays apparatus was used to correlate counting rates measured with the {sup 6}LiI(Eu) detection system and doses measured with an ionization chamber at several distances from the X ray source. (author)

  1. Measurement of gamma-ray production cross sections in neutron-induced reactions for Al and Pb

    International Nuclear Information System (INIS)

    Pavlik, A.; Vonach, H.; Hitzenberger, H.

    1995-01-01

    The prompt gamma-radiation from the interaction of fast neutrons with aluminum and lead was measured using the white neutron beam of the WNR facility at the Los Alamos National Laboratory. The samples (Al and isotopically enriched 207 Pb and 208 Pb) were positioned at about 20 m or 41 m distance from the neutron production target. The spectra of the emitted gamma-rays were measured with a high-resolution HPGe detector. The incident neutron energy was determined by the time-of-flight method and the neutron fluence was measured with a U fission chamber. From the aluminum gamma-ray spectra excitation functions for prominent gamma-transitions in various residual nuclei (in the range from O to Al) were derived for neutron energies from 3 MeV to 400 MeV. For lead (n,xnγ) reactions were studied for neutron energies up to 200 MeV by analyzing prominent gamma-transitions in the residual nuclei 200,202,204,206,207,208 Pb. The experimental results were compared with nuclear model calculations using the code GNASH. A good overall agreement was obtained without special parameter adjustments

  2. Relative biological effectiveness (RBE) of fission neutrons and gamma rays at occupational exposure levels: Volume 1, Studies on the genetic effects in mice of 60 equal once-weekly exposures to fission neutrons and gamma rays

    International Nuclear Information System (INIS)

    Grahn, D.; Carnes, B.A.

    1987-10-01

    The relative biological effectiveness (RBE) values for low doses of fission neutrons compared to 60 Co gamma rays were determined with four separate assessments of genetic damage induced in young hybrid male mice. Both radiations were delivered at low dose levels over about one-half the adult lifetime as 60 once-weekly exposures. Genetic damage assessed included both transient and residual injury. The latter is more critical, as residual genetic injury can be transmitted to subsequent generations long after the radiation exposures have ceased. Assays were performed periodically during the 60-week exposure period and at 10 or more weeks after the irradiations had terminated. RBE values, with few exceptions, ranged between 5 and 15 for transient injury and between 25 and 50 for different types of residual genetic injury. The most important form of residual genetic damage in this study was the balanced reciprocal chromosome translocation. These translocations continue to be transmitted throughout reproductive life and can lead to reduced fertility and increased prenatal mortality. The best estimate of the RBE value for translocations was 45 +- 10. Implications and recommendations with regard to the neutron quality factor will be presented conjointly with the findings from the data obtained in this same project on life shortening and on the risks of incidence or death from neoplastic disease. 64 refs., 23 tabs

  3. Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

    International Nuclear Information System (INIS)

    Kwon, S.G.; Lee, S.Y.; Yook, C.C.

    1981-01-01

    This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

  4. Prospects for joint observations of gravitational waves and gamma rays from merging neutron star binaries

    Energy Technology Data Exchange (ETDEWEB)

    Patricelli, B.; Razzano, M.; Fidecaro, F. [Dipartimento di Fisica, Università di Pisa, Largo B. Pontecorvo, 3, 56127 Pisa (Italy); Cella, G. [INFN—Sezione di Pisa, Largo B. Pontecorvo, 3, 56127 Pisa (Italy); Pian, E.; Stamerra, A. [Scuola Normale Superiore, Piazza dei Cavalieri, 7, 56126 Pisa (Italy); Branchesi, M., E-mail: barbara.patricelli@pi.infn.it, E-mail: massimiliano.razzano@unipi.it, E-mail: giancarlo.cella@pi.infn.it, E-mail: francesco.fidecaro@unipi.it, E-mail: elena.pian@sns.it, E-mail: marica.branchesi@uniurb.it, E-mail: stamerra@oato.inaf.it [Universit\\a di Urbino, Via Aurelio Saffi, 2, 61029 Urbino (Italy)

    2016-11-01

    The detection of the events GW150914 and GW151226, both consistent with the merger of a binary black hole system (BBH), opened the era of gravitational wave (GW) astronomy. Besides BBHs, the most promising GW sources are the coalescences of binary systems formed by two neutron stars or a neutron star and a black hole. These mergers are thought to be connected with short Gamma Ray Bursts (GRBs), therefore combined observations of GW and electromagnetic (EM) signals could definitively probe this association. We present a detailed study on the expectations for joint GW and high-energy EM observations of coalescences of binary systems of neutron stars with Advanced Virgo and LIGO and with the Fermi gamma-ray telescope. To this scope, we designed a dedicated Montecarlo simulation pipeline for the multimessenger emission and detection by GW and gamma-ray instruments, considering the evolution of the GW detector sensitivities. We show that the expected rate of joint detection is low during the Advanced Virgo and Advanced LIGO 2016–2017 run; however, as the interferometers approach their final design sensitivities, the rate will increase by ∼ a factor of ten. Future joint observations will help to constrain the association between short GRBs and binary systems and to solve the puzzle of the progenitors of GWs. Comparison of the joint detection rate with the ones predicted in this paper will help to constrain the geometry of the GRB jet.

  5. Radiography apparatus using gamma rays emitted by water activated by fusion neutrons

    Science.gov (United States)

    Smith, Donald L.; Ikeda, Yujiro; Uno, Yoshitomo

    1996-01-01

    Radiography apparatus includes an arrangement for circulating pure water continuously between a location adjacent a source of energetic neutrons, such as a tritium target irradiated by a deuteron beam, and a remote location where radiographic analysis is conducted. Oxygen in the pure water is activated via the .sup.16 O(n,p).sup.16 N reaction using .sup.14 -MeV neutrons produced at the neutron source via the .sup.3 H(d,n).sup.4 He reaction. Essentially monoenergetic gamma rays at 6.129 (predominantly) and 7.115 MeV are produced by the 7.13-second .sup.16 N decay for use in radiographic analysis. The gamma rays have substantial penetrating power and are useful in determining the thickness of materials and elemental compositions, particularly for metals and high-atomic number materials. The characteristic decay half life of 7.13 seconds of the activated oxygen is sufficient to permit gamma ray generation at a remote location where the activated water is transported, while not presenting a chemical or radioactivity hazard because the radioactivity falls to negligible levels after 1-2 minutes.

  6. Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4

    International Nuclear Information System (INIS)

    Baur, A.; Dupont, C.; Totth, B.

    1978-01-01

    The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks

  7. Polyethylene-reflected plutonium metal sphere : subcritical neutron and gamma measurements.

    Energy Technology Data Exchange (ETDEWEB)

    Mattingly, John K.

    2009-11-01

    Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.

  8. Polyethylene-reflected plutonium metal sphere: subcritical neutron and gamma measurements

    International Nuclear Information System (INIS)

    Mattingly, John K.

    2009-01-01

    Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.

  9. Evaluation of an automated assay system to measure soil radionuclides by L x-ray and gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Nyhan, J.W.; Drennon, B.J.; Crowell, J.M.

    1982-08-01

    An automated radionuclide assay system for conducting soil radioassays using L x-ray and gamma-ray spectrometry was evaluated. Wet chemistry assay procedures were shown to be considerably more time consuming than similar analyses of soil on this radionuclide assay system. The detection limits of 241 Am and plutonium were determined, as well as the reproducibility of radionuclide assay results. The L x-ray spectrometric measurements were compared with radiochemical analyses on several tuff samples. The assay system's intrinsic germanium detector was found to respond linearly to varying low concentrations of 241 Am and plutonium, both of which were easily detected in the presence of elevated concentrations of 137 Cs

  10. Can a large neutron excess help solve the baryon loading problem in gamma-Ray burst fireballs?

    Science.gov (United States)

    Fuller; Pruet; Abazajian

    2000-09-25

    We point out that the baryon loading problem in gamma-ray burst (GRB) models can be ameliorated if a significant fraction of the baryons which inertially confine the fireball is converted to neutrons. A high neutron fraction can result in a reduced transfer of energy from relativistic light particles in the fireball to baryons. The energy needed to produce the required relativistic flow in the GRB is consequently reduced, in some cases by orders of magnitude. A high neutron-to-proton ratio has been calculated in neutron star-merger fireball environments. Significant neutron excess also could occur near compact objects with high neutrino fluxes.

  11. Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hausladen, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Newby, Jason [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-11-01

    The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the 235U content of the fuel with a moderated 241Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the

  12. A real-time neutron-gamma discriminator based on the support vector machine method for the time-of-flight neutron spectrometer

    Science.gov (United States)

    Wei, ZHANG; Tongyu, WU; Bowen, ZHENG; Shiping, LI; Yipo, ZHANG; Zejie, YIN

    2018-04-01

    A new neutron-gamma discriminator based on the support vector machine (SVM) method is proposed to improve the performance of the time-of-flight neutron spectrometer. The neutron detector is an EJ-299-33 plastic scintillator with pulse-shape discrimination (PSD) property. The SVM algorithm is implemented in field programmable gate array (FPGA) to carry out the real-time sifting of neutrons in neutron-gamma mixed radiation fields. This study compares the ability of the pulse gradient analysis method and the SVM method. The results show that this SVM discriminator can provide a better discrimination accuracy of 99.1%. The accuracy and performance of the SVM discriminator based on FPGA have been evaluated in the experiments. It can get a figure of merit of 1.30.

  13. Estimated Uncertainty in Segmented Gamma Scanner Assay Results due to the Variation in Drum Tare Weights

    International Nuclear Information System (INIS)

    Bosko, A.; Croft, St.; Gulbransen, E.

    2009-01-01

    General purpose gamma scanners are often used to assay unknown drums that differ from those used to create the default calibration. This introduces a potential source of bias into the matrix correction when the correction is based on the estimation of the mean density of the drum contents from a weigh scale measurement. In this paper we evaluate the magnitude of this bias that may be introduced by performing assay measurements with a system whose matrix correction algorithm was calibrated with a set of standard drums but applied to a population of drums whose tare weight may be different. The matrix correction factors are perturbed in such cases because the unknown difference in tare weight gets reflected as a bias in the derived matrix density. This would be the only impact if the difference in tare weight was due solely to the weight of the lid or base, say. But in reality the reason for the difference may be because the steel wall of the drum is of a different thickness. Thus, there is an opposing interplay at work which tends to compensate. The purpose of this work is to evaluate and bound the magnitude of the resulting assay uncertainty introduced by tare weight variation. We compare the results obtained using simple analytical models and the 3-D ray tracing with ISOCS software to illustrate and quantify the problem. The numerical results allow a contribution to the Total Measurement Uncertainty (TMU) to be propagated into the final assay result. (authors)

  14. Neutron and gamma probes: Their use in agronomy

    International Nuclear Information System (INIS)

    Bacchi, O.O.; Reichart, K.; Calvache, M.

    2002-01-01

    The concept of this training manual originated during a regional training workshop on the use of neutron probe in water and nutrient balance studies, organized in 1997 in the frame of an IAEA Regional Technical Co-operation Project for Latin America entitle Plant Nutrition, Water and Soil Management, for which the integrated approach was adopted. The original version (in Spanish) was a comprehensive manual covering theoretical and practical aspects required for the proper utilization of the equipment. The contributions of the peer reviewers, editors and technical translators of the three versions in English, French and Spanish have greatly enhanced the content and quality of the manual. It is hoped that this manual will be useful for future training events and serve as a key reference to soil/water scientists in the field of sustainable management of scarce water resources in both rain-fed and irrigated agricultural production systems

  15. Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)

    2009-08-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.

  16. Pulse-shape discrimination of high-energy neutrons and gamma rays in NaI(Tl)

    International Nuclear Information System (INIS)

    Share, G.H.; Kurfess, J.D.; Theus, R.B.

    1978-01-01

    Pulse-shape discrimination can be used to separate neutron and gamma-ray interactions depositing energies up to in excess of 50 MeV in NaI(Tl) crystals. The secondary alpha particles, deuterons and protons produced in the neutron interactions are also resolvable. (Auth.)

  17. Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

    Science.gov (United States)

    Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

    2018-01-01

    University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

  18. Neutron-gamma discrimination via PSD plastic scintillator and SiPMs

    Science.gov (United States)

    Taggart, M. P.; Payne, C.; Sellin, P. J.

    2016-10-01

    The reduction in availability and inevitable increase in cost of traditional neutron detectors based on the 3He neutron capture reaction has resulted in a concerted effort to seek out new techniques and detection media to meet the needs of national nuclear security. Traditionally, the alternative has been provided through pulse shape discrimination (PSD) using liquid scintillators. However, these are not without their own inherent issues, primarily concerning user safety and ongoing maintenance. A potential system devised to separate neutron and gamma ray pulses utilising the PSD technique takes advantage of recent improvements in silicon photomultiplier (SiPM) technology and the development of plastic scintillators exhibiting the PSD phenomena. In this paper we present the current iteration of this ongoing work having achieved a Figure of Merit (FoM) of 1.39 at 1.5 MeVee.

  19. Bulk - Samples gamma-rays activation analysis (PGNAA) with Isotopic Neutron Sources

    International Nuclear Information System (INIS)

    HASSAN, A.M.

    2009-01-01

    An overview is given on research towards the Prompt Gamma-ray Neutron Activation Analysis (PGNAA) of bulk-samples. Some aspects in bulk-sample PGNAA are discussed, where irradiation by isotopic neutron sources is used mostly for in-situ or on-line analysis. The research was carried out in a comparative and/or qualitative way or by using a prior knowledge about the sample material. Sometimes we need to use the assumption that the mass fractions of all determined elements add up to 1. The sensitivity curves are also used for some elements in such complex samples, just to estimate the exact percentage concentration values. The uses of 252 Cf, 241 Arn/Be and 239 Pu/Be isotopic neutron sources for elemental investigation of: hematite, ilmenite, coal, petroleum, edible oils, phosphates and pollutant lake water samples have been mentioned.

  20. Measurement of neutron and gamma absorbed doses in phantoms exposed to mixed fields

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.; Lemaire, G.; Maas, J.

    1985-01-01

    In order to study the dosimetric characteristics of PIN junctions, the absorbed doses measured by junctions and FLi7 in air and water phantoms were compared with the doses measured by classical neutron dosimetry in mixed fields. The validity of the experimental responses of PIN junctions being thus checked and established, neutron and gamma dose distributions in tissue equivalent plastic phantoms (plastinaut) and mammals (piglets) were evaluated as well as the absorbed dose distributions in the pig bone-marrow producing areas. By using correlatively a Monte-Carlo calculation method and applying some simplifying assumptions, the absorbed doses were derived from the spectrum of SILENE's neutrons at various depths inside a cubic water phantom and the results were compared with some from the literature [fr

  1. Neutron and gamma ray calculation for Hiroshima-type atomic bomb

    Energy Technology Data Exchange (ETDEWEB)

    Hoshi, Masaharu; Endo, Satoru; Takada, Jun [Hiroshima Univ. (Japan). Research Inst. for Radiation Biology and Medicine; Iwatani, Kazuo; Oka, Takamitsu; Shizuma, Kiyoshi; Fujita, Shoichiro; Hasai, Hiromi

    1998-03-01

    We looked at the radiation dose of Hiroshima and Nagasaki atomic bomb again in 1986. We gave it the name of ``Dosimetry System 1986`` (DS86). We and other groups have measured the expose dose since 1986. Now, the difference between data of {sup 152}Eu and the calculation result on the basis of DS86 was found. To investigate the reason, we carried out the calculations of neutron transport and neutron absorption gamma ray for Hiroshima atomic bomb by MCNP3A and MCNP4A code. The problems caused by fast neutron {sup 32}P from sulfur in insulator of pole. To correct the difference, we investigated many models and found agreement of all data within 1 km. (S.Y.)

  2. Use of integral experiments to improve neutron propagation and gamma heating calculations

    International Nuclear Information System (INIS)

    Oceraies, Y.; Caumette, P.; Devillers, C.; Bussac, J.

    1979-01-01

    1) The studies to define and improve the accuracies of neutron propagation and gamma heating calculations from integral experiments are encompassed in the field of the fast reactor physics program at CEA. 2) A systematic analysis of neutron propagation in Fe-Na clean media, with variable volumic composition between 0 and 100% in sodium, has been performed on the HARMONIE source reactor. Gamma heating traverses in the core, the blankets and several control rods, have been measured in the R Z core program at MASURCA. The experimental techniques, the accuracies and the results obtained are given. The approximations of the calculational methods used to analyse these experiments and to predict the corresponding design parameters are also described. 3) Particular emphasis is given to the methods planned to improve fundamental data used in neutron propagation calculations, using the discrepancies observed between measured and calculated results in clean integral experiments. One of these approaches, similar to the techniques used in core physics, relies upon sensitivity studies and eventually on adjustment techniques applied to neutron propagation. (author)

  3. Method and apparatus for neutron induced gamma ray logging for lithology identificaion

    International Nuclear Information System (INIS)

    Oliver, D.W.; Culver, R.B.

    1979-01-01

    A pulsed neutron generator in a well logging instrument is pulsed at a clock frequency of 20 KHz. Inelastic scatter gamma rays are detected during a first time interval coinciding with the neutron source being on and capture gamma rays are measured during a second interval subsequent to the end of each neutron burst. Only a single detected pulse, assuming detection occurs, is transmitted during each of the two detection intervals. Sync pulses are generated in the well logging instrument scaled down to a frequency of 200 Hz for transmission to the earth's surface. At the earth's surface, the scaled-down sync pulses are applied to a phase-locked loop system for regenerating the sync pulses to the same frequency as that of the clock frequency used to pulse the neutron source and to open the detection gates in the borehole instrument. The regenerated sync pulses are used in the surface instrumentation to route the pulses occurring in the inelastic interval into one section of a multichannel analyzer memory and the pulses occurring in the capture interval into another section of the multichannel analyzer. The use of memory address decoders, subtractors and ratio circuits enables both a carbon/oxygen ratio and a silicon/calcium ratio to be struck, substantially independent of the chlorine content of the borehole and formation

  4. Monitoring of processes with gamma-rays of neutron capture and short-living radionuclides

    International Nuclear Information System (INIS)

    Aripov, G.A.; Kurbanov, B.I.; Allamuratova, G.

    2004-01-01

    Element content is a fundamental parameter of a substance, on which all its properties, and also character of physical, chemical, biological, technological and ecological processes depend. Therefore monitoring of element content (in the course of technological process - on line; in natural conditions - in site; or in living organisms - in vivo) becomes necessary for investigation of aforementioned processes. This problem can be successfully solved by using the methods of prompt gamma activation analysis (PGAA) and instrumental neutron activation analysis (INAA) on short-living radionuclides. These methods don't depend on type of substance (biological, geological, technological etc.), since the content is determined by gamma radiation of nuclei, and allows to meet such a serious requirement like the necessity of achieving minimal irradiation of the object and its minimal residual activity. In this work minimal determinable concentrations of various elements are estimated (based on experimental data) by the method of PGAA using radionuclide 252 Cf - source of neutrons with the yield of the oil of 10 8 neutron/sec on the experimental device with preliminary focusing of neutrons /1/, and also data of determination of elements by their isotopes with maximum time efficiency /2,3/ by the method of INAA. (author)

  5. Gamma ray scanner systems for nondestructive assay of heterogeneous waste barrels

    International Nuclear Information System (INIS)

    Martz, H.E.; Decman, B.J.; Roberson, G.P.; Levai, F.

    1997-01-01

    Traditional gamma safeguards measurements have usually been performed using a segmented gamma scanning (SGS) system. The accuracy of this technique relies on the assumption that the sample matrix and the activity are both uniform for a segment. Waste barrels are often highly heterogeneous, span a wide range of composition and matrix type. The primary sources of error are all directly or indirectly related to a non-uniform measurement response associated with unknown radioactive source spatial distribution and heterogeneity of the matrix. These errors can be significantly reduced by some imaging techniques that measure exact spatial locations of sources and attenuation maps. In this paper we describe a joint R ampersand D effort between the Lawrence Livermore National Laboratory (LLNL) and the Institute of Nuclear Techniques (INT) of the Technical University, Budapest, to compare results obtained by two different gamma-ray nondestructive assay (NDA) systems used for imaging waste barrels. The basic principles are the same, but the approaches are different. Key factors to judge the adequacy of a method are the detection limit and the accuracy. Test drums representing waste to be measured are used to determine basic parameters of these techniques

  6. A weighted least-squares lump correction algorithm for transmission-corrected gamma-ray nondestructive assay

    International Nuclear Information System (INIS)

    Prettyman, T.H.; Sprinkle, J.K. Jr.; Sheppard, G.A.

    1993-01-01

    With transmission-corrected gamma-ray nondestructive assay instruments such as the Segmented Gamma Scanner (SGS) and the Tomographic Gamma Scanner (TGS) that is currently under development at Los Alamos National Laboratory, the amount of gamma-ray emitting material can be underestimated for samples in which the emitting material consists of particles or lumps of highly attenuating material. This problem is encountered in the assay of uranium and plutonium-bearing samples. To correct for this source of bias, we have developed a least-squares algorithm that uses transmission-corrected assay results for several emitted energies and a weighting function to account for statistical uncertainties in the assay results. The variation of effective lump size in the fitted model is parameterized; this allows the correction to be performed for a wide range of lump-size distributions. It may be possible to use the reduced chi-squared value obtained in the fit to identify samples in which assay assumptions have been violated. We found that the algorithm significantly reduced bias in simulated assays and improved SGS assay results for plutonium-bearing samples. Further testing will be conducted with the TGS, which is expected to be less susceptible than the SGS to systematic source of bias

  7. Prompt gamma neutron activation analysis facility at the RA-6 research reactor

    International Nuclear Information System (INIS)

    Sanchez, F. A.; Calzetta, O

    2004-01-01

    A prompt gamma neutron activation activation analysis facility was developed at the 500 kw thermal power RA-6 research reactor of the Bariloche Atomic Center, Argentina.This facility consist of a radial beam port with external positioning of the sample.The gamma radiation is reduced by a bismuth filter placed inside the extraction tube and the beam diameter is limited by a set of two collimators up to 5 cm.The neutron flux at the sample position is 7 10 6 n/cm 2 s with a Cadmium ratio of 20/1.The gamma detector is a 50 % efficiency type p HPGe rounded by a NaI(Tl) for Compton suppressioning.The gamma spectra is measured through 0 to 8.5 MeV.The background have counting rate of 350 cps without sample. In this work is shown the efficiency curve, the calculed sensibilities and the lower detection limits for B, Cd, Sm, Gd, H, Cl, Hg, Eu, Ti, Ag, Au, Mo. The RA-6's PGNAA facility is fully working, although the analytic capacity is under improvement [es

  8. Neutron-capture gamma-ray analysis of coal for sulfur, iron, silicon and moisture

    International Nuclear Information System (INIS)

    Fay, D.A.

    1979-05-01

    Samples of coal weighing approximately 200 grams placed in a collimated beam of neutrons from the thermal column of the Ames Laboratory Research Reactor produced capture gamma-rays which could be used for the simultaneous determination of sulfur and iron. Spectra from NaI(Tl) and Ge(Li) detectors were used and interferences were located by examining spectra of the major elemental components of coal. In determining sulfur, iron is a potential source of interference when gamma-ray spectra are collected with a NaI(Tl) detector. Corrections for iron interference were made by use of a higher energy iron peak. The possibility of determining silicon in coal was investigated but this element determination was unsuccessful since capture gamma-ray spectrometry lacked the necessary sensitivity for silicon. A linear relation was found between the area of the hydrogen capture peak at 2.23 MeV and the amount of water added to coal

  9. Gamma-ray scanning of neutron activated geological sediments for studying elemental profile distributions

    International Nuclear Information System (INIS)

    Ellinger, M.; Janghorbani, M.; Starke, K.

    1976-01-01

    Gamma-ray scanning for application to elemental profile studies of geological samples was studied with a neutron activated Baltic Shield sediment. Profile distribution of seven elements were measured. The capabilities and limitations of gamma-ray scanning are discussed by comparing the results with profiles obtained after the mechanical subdivision of the sample and the activation of the appropriately sized separates. With respect to the merits and limitations of scanning gamma-ray spectrometry applied to activated complex matrices the following conclusions were drawn. Qualitatively, the scanning method yields the same information as the much more laborious method of mechanical sudbisubdivision. Quantitatively, it is significantly less accurate. The scanning method has the significant advantage of allowing preservation of the sample. This could be important for such speciments as lunar and archeological materials. The method reduces sample preparation time and the possibility of sample contamination. (T.G.)

  10. Whole blood interferon-gamma assay for baseline tuberculosis screening among Japanese healthcare students.

    Directory of Open Access Journals (Sweden)

    Katsuyuki Hotta

    Full Text Available BACKGROUND: The whole blood interferon-gamma assay (QuantiFERON-TB-2G; QFT has not been fully evaluated as a baseline tuberculosis screening test in Japanese healthcare students commencing clinical contact. The aim of this study was to compare the results from the QFT with those from the tuberculin skin test (TST in a population deemed to be at a low risk for infection with Mycobacterium tuberculosis. METHODOLOGY/PRINCIPAL FINDINGS: Healthcare students recruited at Okayama University received both the TST and the QFT to assess the level of agreement between these two tests. The interleukin-10 levels before and after exposure to M tuberculosis-specific antigens (early-secreted antigenic target 6-kDa protein [ESAT-6] and culture filtrate protein 10 [CFP-10] were also measured. Of the 536 healthcare students, most of whom had been vaccinated with bacillus-Calmette-Guérin (BCG, 207 (56% were enrolled in this study. The agreement between the QFT and the TST results was poor, with positive result rates of 1.4% vs. 27.5%, respectively. A multivariate analysis also revealed that the induration diameter of the TST was not affected by the interferon-gamma concentration after exposure to either of the antigens but was influenced by the number of BCG needle scars (p = 0.046. The whole blood interleukin-10 assay revealed that after antigen exposure, the median increases in interleukin-10 concentration was higher in the subgroup with the small increase in interferon-gamma concentration than in the subgroup with the large increase in interferon-gamma concentration (0.3 vs. 0 pg/mL; p = 0.004. CONCLUSIONS/SIGNIFICANCE: As a baseline screening test for low-risk Japanese healthcare students at their course entry, QFT yielded quite discordant results, compared with the TST, probably because of the low specificity of the TST results in the BCG-vaccinated population. We also found, for the first time, that the change in the interleukin-10 level after exposure to

  11. Measurement of secondary neutrons and gamma rays produced by neutron interactions in aluminum over the incident energy range 1 to 20 MeV

    International Nuclear Information System (INIS)

    Morgan, G.L.

    1975-11-01

    The spectra of secondary neutrons and gamma rays produced by neutron interaction in a thin sample (approximately 1/6 mean free path) of aluminum have been measured as a function of the incident neutron energy over the range 1 to 20 MeV. Data were taken at an angle of 125 0 . A linac (ORELA) was used as a neutron source with a 47-m flight path. Incident energy was determined by time-of-flight, while secondary spectra were determined by pulse-height unfolding techniques. The results of the measurements are presented in forms suitable for comparison to calculations based on the evaluated data files. (6 tables, 4 figures)

  12. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Directory of Open Access Journals (Sweden)

    Kooyman Timothée

    2017-01-01

    Full Text Available Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  13. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Science.gov (United States)

    Kooymana, Timothée; Buiron, Laurent; Rimpault, Gérald

    2017-09-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long and short term neutron and gamma source is carried out while in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  14. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    International Nuclear Information System (INIS)

    Kooyman, T.; Buiron, L.; Rimpault, G.

    2017-01-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing. (authors)

  15. Gamma-ray imaging and holdup assays of 235-F PuFF cells 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    Aucott, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-12-20

    Savannah River National Laboratory (SRNL) Nuclear Measurements (L4120) was tasked with performing enhanced characterization of the holdup in the PuFF shielded cells. Assays were performed in accordance with L16.1-ADS-2460 using two high-resolution gamma-ray detectors. The first detector, an In Situ Object Counting System (ISOCS)-characterized detector, was used in conjunction with the ISOCS Geometry Composer software to quantify grams of holdup. The second detector, a Germanium Gamma-ray Imager (GeGI), was used to visualize the location and relative intensity of the holdup in the cells. Carts and collimators were specially designed to perform optimum assays of the cells. Thick, pencil-beam tungsten collimators were fabricated to allow for extremely precise targeting of items of interest inside the cells. Carts were designed with a wide range of motion to position and align the detectors. A total of 24 measurements were made, each typically 24 hours or longer to provide sufficient statistical precision. This report presents the results of the enhanced characterization for cells 1 and 2. The measured gram values agree very well with results from the 2014 study. In addition, images were created using both the 2014 data and the new GeGI data. The GeGI images of the cells walls reveal significant Pu-238 holdup on the surface of the walls in cells 1 and 2. Additionally, holdup is visible in the two pass-throughs from cell 1 to the wing cabinets. This report documents the final element (exterior measurements coupled with gamma-ray imaging and modeling) of the enhanced characterization of cells 1-5 (East Cell Line).

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Fast neutron-gamma discrimination on neutron emission profile measurement on JT-60U

    International Nuclear Information System (INIS)

    Ishii, K.; Okamoto, A.; Kitajima, S.; Sasao, M.; Shinohara, K.; Ishikawa, M.; Baba, M.; Isobe, M.

    2010-01-01

    A digital signal processing (DSP) system is applied to stilbene scintillation detectors of the multichannel neutron emission profile monitor in JT-60U. Automatic analysis of the neutron-γ pulse shape discrimination is a key issue to diminish the processing time in the DSP system, and it has been applied using the two-dimensional (2D) map. Linear discriminant function is used to determine the dividing line between neutron events and γ-ray events on a 2D map. In order to verify the validity of the dividing line determination, the pulse shape discrimination quality is evaluated. As a result, the γ-ray contamination in most of the beam heating phase was negligible compared with the statistical error with 10 ms time resolution.

  18. A Gamma Polarimeter for Neutron Polarization Measurement in a Liquid Deuterium Target for Parity Violation in Polarized Neutron Capture on Deuterium.

    Science.gov (United States)

    Komives, A; Sint, A K; Bowers, M; Snow, M

    2005-01-01

    A measurement of the parity-violating gamma asymmetry in n-D capture would yield information on N-N parity violation independent of the n-p system. Since cold neutrons will depolarize in a liquid deuterium target in which the scattering cross section is much larger than the absorption cross section, it will be necessary to quantify the loss of polarization before capture. One way to do this is to use the large circular polarization of the gamma from n-D capture and analyze the circular polarization of the gamma in a gamma polarimeter. We describe the design of this polarimeter.

  19. Unsteady Plasma Ejections from Hollow Accretion Columns of Galactic Neutron Stars as a Trigger for Gamma-Ray Bursts

    Science.gov (United States)

    Gvaramadze, V. V.

    1995-09-01

    We propose a model of gamma-ray bursts (GRBs) based on close Galactic neutron stars with accretion disks. We outline a simple mechanism of unsteady plasma ejections during episodic accretion events. The relative kinetic energy of ejected blobs can be converted into gamma-rays by internal shocks. The beaming of gamma-ray emission can be responsible for the observed isotropic angular distribution of GRBs.

  20. Sample design and gamma-ray counting strategy of neutron activation system for triton burnup measurements in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jungmin [Department of Energy System Engineering, Seoul National University, Seoul (Korea, Republic of); Cheon, Mun Seong [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Chung, Kyoung-Jae, E-mail: jkjlsh1@snu.ac.kr [Department of Energy System Engineering, Seoul National University, Seoul (Korea, Republic of); Hwang, Y.S. [Department of Energy System Engineering, Seoul National University, Seoul (Korea, Republic of)

    2016-11-01

    Highlights: • Sample design for triton burnup ratio measurement is carried out. • Samples for 14.1 MeV neutron measurements are selected for KSTAR. • Si and Cu are the most suitable materials for d-t neutron measurements. • Appropriate γ-ray counting strategies for each selected sample are established. - Abstract: On the purpose of triton burnup measurements in Korea Superconducting Tokamak Advanced Research (KSTAR) deuterium plasmas, appropriate neutron activation system (NAS) samples for 14.1 MeV d-t neutron measurements have been designed and gamma-ray counting strategy is established. Neutronics calculations are performed with the MCNP5 neutron transport code for the KSTAR neutral beam heated deuterium plasma discharges. Based on those calculations and the assumed d-t neutron yield, the activities induced by d-t neutrons are estimated with the inventory code FISPACT-2007 for candidate sample materials: Si, Cu, Al, Fe, Nb, Co, Ti, and Ni. It is found that Si, Cu, Al, and Fe are suitable for the KSATR NAS in terms of the minimum detectable activity (MDA) calculated based on the standard deviation of blank measurements. Considering background gamma-rays radiated from surrounding structures activated by thermalized fusion neutrons, appropriate gamma-ray counting strategy for each selected sample is established.