WorldWideScience

Sample records for neutron absorber techniques

  1. Methods for absorbing neutrons

    Science.gov (United States)

    Guillen, Donna P [Idaho Falls, ID; Longhurst, Glen R [Idaho Falls, ID; Porter, Douglas L [Idaho Falls, ID; Parry, James R [Idaho Falls, ID

    2012-07-24

    A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.

  2. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1983-01-01

    A neutron-absorber body for use in burnable poison rods in a nuclear reactor. The body is composed of a matrix of Al 2 O 3 containing B 4 C, the neutron absorber. Areas of high density polycrystalline Al 2 O 3 particles are predominantly encircled by pores in some of which there are B 4 C particles. This body is produced by initially spray drying a slurry of A1 2 O 3 powder to which a binder has been added. The powder of agglomerated spheres of the A1 2 O 3 with the binder are dry mixed with B 4 C powder. The mixed powder is formed into a green body by isostatic pressure and the green body is sintered. The sintered body is processed to form the neutron-absorber body. In this case the B 4 C particles are separate from the spheres resulting from the spray drying instead of being embedded in the sphere

  3. Neutron absorbing article

    International Nuclear Information System (INIS)

    Naum, R.G.; Owens, D.P.; Dooker, G.I.

    1981-01-01

    A neutron-absorbing article suitable for use in spent fuel racks is described. It comprises boron carbide particles, diluent particles, and a phenolic polymer cured to a continuous matrix. The diluent may be silicon carbide, graphite, amorphous carbon, alumina, or silica. The combined boron carbide-diluent phase contains no more than 2 percent B 2 O 3 , and the neutron-absorbing article contains from 20 to 40 percent phenol resin. The ratio of boron carbide to diluent particles is in the range 1:9 to 9:1

  4. Neutron absorbing article

    International Nuclear Information System (INIS)

    Naum, R.G.; Owens, D.P.; Dooher, G.I.

    1979-01-01

    A neutron absorbing article, in flat plate form and suitable for use in a storage rack for spent fuel, includes boron carbide particles, diluent particles and a solid, irreversibly cured phenolic polymer cured to a continuous matrix binding the boron carbide and diluent particles. The total conent of boron carbide and diluent particles is a major proportion of the article and the content of cured phenolic polymer present is a minor proportion. By regulation of the ratio of boron carbide particles to diluent particles, normally within the range of 1:9 and 9:1 and preferably within the range of 1:5 to 5:1, the neutron absorbing activity of the product may be controlled, which facilitates the manufacture of articles of particular absorbing activities best suitable for specific applications

  5. Corrosion resistant neutron absorbing coatings

    Science.gov (United States)

    Choi, Jor-Shan [El Cerrito, CA; Farmer, Joseph C [Tracy, CA; Lee, Chuck K [Hayward, CA; Walker, Jeffrey [Gaithersburg, MD; Russell, Paige [Las Vegas, NV; Kirkwood, Jon [Saint Leonard, MD; Yang, Nancy [Lafayette, CA; Champagne, Victor [Oxford, PA

    2012-05-29

    A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.

  6. Quantitative neutron radiography using neutron absorbing honeycomb

    International Nuclear Information System (INIS)

    Tamaki, Masayoshi; Oda, Masahiro; Takahashi, Kenji; Ohkubo, Kohei; Tasaka, Kanji; Tsuruno, Akira; Matsubayashi, Masahito.

    1993-01-01

    This investigation concerns quantitative neutron radiography and computed tomography by using a neutron absorbing honeycomb collimator. By setting the neutron absorbing honeycomb collimator between object and imaging system, neutrons scattered in the object were absorbed by the honeycomb material and eliminated before coming to the imaging system, but the neutrons which were transmitted the object without interaction could reach the imaging system. The image by purely transmitted neutrons gives the quantitative information. Two honeycombs were prepared with coating of boron nitride and gadolinium oxide and evaluated for the quantitative application. The relation between the neutron total cross section and the attenuation coefficient confirmed that they were in a fairly good agreement. Application to quantitative computed tomography was also successfully conducted. The new neutron radiography method using the neutron-absorbing honeycomb collimator for the elimination of the scattered neutrons improved remarkably the quantitativeness of the neutron radiography and computed tomography. (author)

  7. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1985-01-01

    This patent deals with the fabrication of pellets for neutron absorber rods. Such a pellet includes a matrix of a refractory material which may be aluminum or zirconium oxide, and a burnable poison distributed throughout the matrix. The neutron absorber material may consist of one or more elements or compounds of the metals boron, gadolinium, samarium, cadmium, europium, hafnium, dysprosium and indium. The method of fabricating pellets of these materials outlined in this patent is designed to produce pores or voids in the pellets that can be used to take up the expansion of the burnable poison and to absorb the helium gas generated. In the practice of this invention a slurry of Al 2 O 3 is produced. A hard binder is added and the slurry and binder are spray dried. This powder is mixed with dry B 4 C powder, forming a homogeneous mixture. This mixture is pressed into green tubes which are then sintered. During sintering the binder volatilizes leaving a ceramic with nearly spherical high-density regions of

  8. Neutron absorbing element

    International Nuclear Information System (INIS)

    Kasai, Shigeo.

    1991-01-01

    The present invention concerns a neutron absorbing element of a neutron shielding member used for an LMFBR type reactor. The inside of a fuel can sealed at both of the upper and the lower ends thereof with plugs is partitioned into an upper and a lower chambers by an intermediate plug. A discharging hole is disposed at the upper end plug, which is in communication with the outside. A communication tube is disposed at the intermediate end plug and it is in communication with the lower chamber containing B 4 C pellets. A cylindrical support member having three porous plugs connected in series is disposed at the lower surface of the discharging hole provided at the upper end plug. Further, the end of the discharging hole is sealed with high temperature solder and He atmosphere is present at the inside of the fuel can. With such a constitution, the supporting differential pressure of the porous plugs can be made greater while discharging He gases generated from B 4 C to the outside. Further, the porous plugs can be surely wetted by coolants. Accordingly, it is possible to increase life time and shorten the size. (I.N.)

  9. Neutron techniques

    International Nuclear Information System (INIS)

    Charlton, J.S.

    1986-01-01

    The way in which neutrons interact with matter such as slowing-down, diffusion, neutron absorption and moderation are described. The use of neutron techniques in industry, in moisture gages, level and interface measurements, the detection of blockages, boron analysis in ore feedstock and industrial radiography are discussed. (author)

  10. Neutron-absorbing alloys

    International Nuclear Information System (INIS)

    Portnoi, K.I.; Arabei, L.B.; Gryaznov, G.M.; Levi, L.I.; Lunin, G.L.; Kozhukhov, V.M.; Markov, J.M.; Fedotov, M.E.

    1975-01-01

    A process is described for the production of an alloy consiting of 1 to 20% In, 0.5 to 15% Sm, and from 3 to 18% Hf, the balance being Ni. Such alloys show a good absorption capacity for thermal and intermediate neutrons, good neutron capture efficiency, and good corrosion resistance, and find application in nuclear reactor automatic control and safety systems. The Hf provides for the maintenance of a reasonably high order of neutron capture efficiency throughout the lifetime of a reactor. The alloys are formed in a vacuum furnace operating with an inert gas atmosphere at 280 to 300 mm.Hg. They have a corrosion resistance from 3 to 3.5 times that of the Ag-based alloys commonly employed, and a neutron capture efficiency about twice that of the Ag alloys. Castability and structural strength are good. (U.K.)

  11. Heterogeneous neutron absorbers development

    International Nuclear Information System (INIS)

    Boccaccini, Aldo; Agueda, Horacio; Russo, Diego; Perez, Edmundo

    1987-01-01

    The use of solid burnable absorber materials in power light water reactors has increased in the last years, specially due to improvements attained in costs of generated electricity. The present work summarizes the basic studies made on an alumina-gadolinia system, where alumina is the inert matrix and gadolinia acts as burnable poison, and describes the fabrication method of pellets with that material. High density compacts were obtained in the range of concentrations used by cold pressing and sintering at 1600 deg C in inert (Ar) atmosphere. Finally, the results of the irradiation experiences made at RA-6 reactor, located at the Bariloche Atomic Center, are given where variations on negative reactivity caused by introduction of burnable poison rods were measured. The results obtained from these experiences are in good agreement with those coming from calculation codes. (Author)

  12. Absorbing rods for nuclear fast neutron reactor absorbing assembly

    International Nuclear Information System (INIS)

    Aji, M.; Ballagny, A.; Haze, R.

    1986-01-01

    The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

  13. Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2014-10-15

    In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al{sup 27}, C{sup 12}, B{sup 11}, B{sup 10} and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B{sup 10} content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B{sup 10} content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B{sup 10} content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B{sup 10} content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10{sup 10} order, however, usual neutron flux from spent fuel is 10{sup 8} order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B{sup 10} content is little decreased, so, initial neutron absorbing ability could be kept continuously.

  14. Development of highly effective neutron shields and neutron absorbing materials

    International Nuclear Information System (INIS)

    Tsuda, K.; Matsuda, F.; Taniuchi, H.; Yuhara, T.; Iida, T.

    1993-01-01

    A wide range of materials, including polymers and hydrogen-occluded alloys that might be usable as the neutron shielding material were examined. And a wide range of materials, including aluminum alloys that might be usable as the neutron-absorbing material were examined. After screening, the candidate material was determined on the basis of evaluation regarding its adaptabilities as a high-performance neutron-shielding and neutron-absorbing material. This candidate material was manufactured for trial, after which material properties tests, neutron-shielding tests and neutron-absorbing tests were carried out on it. The specifications of this material were thus determined. This research has resulted in materials of good performance; a neutron-shielding material based on ethylene propylene rubber and titanium hydride, and a neutron-absorbing material based on aluminum and titanium hydride. (author)

  15. Neutron absorbed dose in a pacemaker CMOS

    International Nuclear Information System (INIS)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L.

    2012-01-01

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10 -17 Gy per neutron emitted by the source. (Author)

  16. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-06-15

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

  17. Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-10-01

    The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

  18. Aluminum alloy excellent in neutron absorbing performance

    International Nuclear Information System (INIS)

    Iida, Tetsuya; Tamamura, Tadao; Morimoto, Hiroyuki; Ouchi, Ken-ichiro.

    1987-01-01

    Purpose: To obtain structural materials made of aluminum alloys having favorable neutron absorbing performance and excellent in the performance as structural materials such as processability and strength. Constitution: Powder of Gd 2 O 3 as a gadolinium compound or metal gadolinium is uniformly mixed with the powder of aluminum or aluminum alloy. The amount of the gadolinium compound added is set to 0.1 - 30 % by weight. No sufficient neutron absorbing performance can be obtained if it is less than 0.1 % by weight, whereas the processability and mechanical property of the alloy are degraded if it exceeds 30 % by weight. Further, the grain size is set to less about 50 μm. Further, since the neutron absorbing performance varies greatly if the aluminum powder size exceeds 100 μm, the diameter is set to less than about 100 μm. These mixtures are molded in a hot press. This enables to obtain aimed structural materials. (Takahashi, M.)

  19. Method for manufacture of neutron absorbing articles

    International Nuclear Information System (INIS)

    Owens, D.

    1980-01-01

    A one-step curing method for the manufacture of a neutron absorbing article which comprises irreversibly curing, in desired article form, a form-retaining mixture of boron carbide particles, curable phenolic resin in solid state and in particula te form and a minor proportion of a liquid medium, which boils at a temperature below 200*c., at an elevated temperature so as to obtain bonding of the irreversibly cured phenolic polymer resulting to the boron carbide particles and production of the neutron absorbing article in desired form

  20. Neutron absorbers and methods of forming at least a portion of a neutron absorber

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna P; Porter, Douglas L; Swank, W David; Erickson, Arnold W

    2014-12-02

    Methods of forming at least a portion of a neutron absorber include combining a first material and a second material to form a compound, reducing the compound into a plurality of particles, mixing the plurality of particles with a third material, and pressing the mixture of the plurality of particles and the third material. One or more components of neutron absorbers may be formed by such methods. Neutron absorbers may include a composite material including an intermetallic compound comprising hafnium aluminide and a matrix material comprising pure aluminum.

  1. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C. Y.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: candy_borja@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    The absorbed dose due to neutrons by a Complementary Metal Oxide Semiconductor (CMOS) has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes a patient that must be treated by radiotherapy with a linear accelerator; the pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. When the Linac is working in Bremsstrahlung mode an undesirable neutron field is produced due to photoneutron reactions; these neutrons could damage the CMOS putting the patient at risk during the radiotherapy treatment. In order to estimate the neutron dose in the CMOS a Monte Carlo calculation was carried out where a full radiotherapy vault room was modeled with a W-made spherical shell in whose center was located the source term of photoneutrons produced by a Linac head operating in Bremsstrahlung mode at 18 MV. In the calculations a phantom made of tissue equivalent was modeled while a beam of photoneutrons was applied on the phantom prostatic region using a field of 10 x 10 cm{sup 2}. During simulation neutrons were isotropically transported from the Linac head to the phantom chest, here a 1 {theta} x 1 cm{sup 2} cylinder made of polystyrene was modeled as the CMOS, where the neutron spectrum and the absorbed dose were estimated. Main damages to CMOS are by protons produced during neutron collisions protective cover made of H-rich materials, here the neutron spectrum that reach the CMOS was calculated showing a small peak around 0.1 MeV and a larger peak in the thermal region, both connected through epithermal neutrons. (Author)

  2. Removing fuelling transient using neutron absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Paquette, S.; Chan, P.K.; Bonin, H.W., E-mail: Stephane.Paquette@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering Dept., Kingston, Ontario (Canada); Pant, A. [Cameco Fuel Manufacturing, Port Hope, Ontario (Canada)

    2012-07-01

    Preliminary criticality and burnup calculation results indicate that by employing a small amount of neutron absorber the fuelling transient, currently occurring in a CANDU 37-element fuel bundle, can be significantly reduced. A parametric study using the Los Alamos National Laboratories' MCNP 5 code and Atomic Energy of Canada Limited's WIMS-AECL 3.1 is presented in this paper. (author)

  3. Neutron absorbers, and the production method

    International Nuclear Information System (INIS)

    Kayano, Hideo; Yajima, Seishi; Oono, Hironori.

    1979-01-01

    Purpose: To integrally sinter a metal powder and a metal network material thereby to obtain a material having a high neutron absorbing function, an excellent corrosion resistance and an excellent oxidation resistance. Method: An element having a high neutron absorbing function, such as Gd, or a compound thereof and a powder of a metal having excellent corrosion resistance, oxidation resistance and ductility, such as Fe, Cr or the like are uniformly mixed with each other. In a case where a substance having a neutron absorbing function is a hydroxide an organic complex or the like, it is formed into a gel-like substance and mixed uniformly with the metal powder, the gel-like substance being pasted, and covered on the surface of the metal powder and dried. Then, the mixture or the dry coated material is extended and the metal network material having excellent corrosion resistance, oxidation resistance and ductility is covered or interposed or between at least one layer of upper, intermediate or lower layers of said laminated material, and thereafter is subjected to cold or hot rolling, and then sintered and furthermore rolled, if necessary, the thus treated material being burned in vacuum or a non-oxidizing atmosphere. (Kamimura, M.)

  4. Neutron techniques in Safeguards

    International Nuclear Information System (INIS)

    Zucker, M.S.

    1982-01-01

    An essential part of Safeguards is the ability to quantitatively and nondestructively assay those materials with special neutron-interactive properties involved in nuclear industrial or military technology. Neutron techniques have furnished most of the important ways of assaying such materials, which is no surprise since the neutronic properties are what characterizes them. The techniques employed rely on a wide selection of the many methods of neutron generation, detection, and data analysis that have been developed for neutron physics and nuclear science in general

  5. Neutron detection technique

    International Nuclear Information System (INIS)

    Oblath, N.S.; Poon, A.W.P.

    2000-01-01

    The Sudbury Neutrino Observatory (SNO) has the ability to measure the total flux of all active flavors of neutrinos using the neutral current reaction, whose signature is a neutron. By comparing the rates of the neutral current reaction to the charged current reaction, which only detects electron neutrinos, one can test the neutrino oscillation hypothesis independent of solar models. It is necessary to understand the neutron detection efficiency of the detector to make use of the neutral current reaction. This report demonstrates a coincidence technique to identify neutrons emitted from the 252 Cf neutron calibration source. The source releases on average four neutrons when a 252 Cf nucleus spontaneously fissions. Each neutron is detected as a separate event when the neutron is captured by a deuteron, releasing a gamma ray of approximately 6.25 MeV. This gamma ray is in turn detected by the photomultiplier tube (PMT) array. By investigating the time and spatial separation between neutron-like events, it is possible to obtain a pure sample of neutrons for calibration study. Preliminary results of the technique applied to two calibration runs are presented

  6. Electrochemical Corrosion Testing of Neutron Absorber Materials

    International Nuclear Information System (INIS)

    Tedd Lister; Ron Mizia; Sandra Birk; Brent Matteson; Hongbo Tian

    2006-01-01

    The Yucca Mountain Project (YMP) has been directed by DOE-RW to develop a new repository waste package design based on the transport, aging, and disposal canister (TAD) system concept. A neutron poison material for fabrication of the internal spent nuclear fuel (SNF) baskets for these canisters needs to be identified. A material that has been used for criticality control in wet and dry storage of spent nuclear fuel is borated stainless steel. These stainless products are available as an ingot metallurgy plate product with a molybdenum addition and a powder metallurgy product that meets the requirements of ASTM A887, Grade A. A new Ni-Cr-Mo-Gd alloy has been developed by the Idaho National Laboratory (INL) with its research partners (Sandia National Laboratory and Lehigh University) with DOE-EM funding provided by the National Spent Nuclear Fuel Program (NSNFP). This neutron absorbing alloy will be used to fabricate the SNF baskets in the DOE standardized canister. The INL has designed the DOE Standardized Spent Nuclear Fuel Canister for the handling, interim storage, transportation, and disposal in the national repository of DOE owned spent nuclear fuel (SNF). A corrosion testing program is required to compare these materials in environmental conditions representative of a breached waste canister. This report will summarize the results of crevice corrosion tests for three alloys in solutions representative of ionic compositions inside the waste package should a breech occur. The three alloys in these tests are Neutronit A978 (ingot metallurgy, hot rolled), Neutrosorb 304B4 Grade A (powder metallurgy, hot rolled), and Ni-Cr-Mo-Gd alloy (ingot metallurgy, hot rolled)

  7. RackSaver neutron absorbing device development and testing

    International Nuclear Information System (INIS)

    Lambert, R.; O'Leary, P.; Roberts, P.

    1996-01-01

    Siemens Power Corporation (SPC), in cooperation with the Electric Power Research Institute (EPRI), has developed the RackSaver neutron absorbing insert. The RackSaver insert can be installed onto spent nuclear fuel assemblies to replace deteriorating Boraflex neutron absorbing material installed in some spent-fuel storage racks. This paper describes results of a development and in-pool demonstration program performed to support potential utilization of the RackSaver neutron absorbing insert by affected utilities. The program objective was to advance the RackSaver concept into a field-demonstrated product. This objective was accomplished through three phases: design, licensing and criticality evaluations, and demonstration testing

  8. Estimate of absorbed dose received by individuals irradiated with neutrons

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Mauricio, C.L.P.

    1995-01-01

    An innovating methodology is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the methodology here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μGy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author). 5 refs., 1 fig., 4 tabs

  9. Fast neutron radiation inactivation of Bacillus subtilis: Absorbed dose determination

    International Nuclear Information System (INIS)

    Song Lingli; Zheng Chun; Ai Zihui; Li Junjie; Dai Shaofeng

    2011-01-01

    In this paper, fast neutron inactivation effects of Bacillus subtilis were investigated with fission fast neutrons from CFBR-II reactor of INPC (Institute of Nuclear Physics and Chemistry) and mono-energetic neutrons from the Van de Graaff accelerator at Peking University. The method for determining the absorbed dose in the Bacillus subtilis suspension contained in test tubes is introduced. The absorbed dose, on account of its dependence on the volume and the form of confined state, was determined by combined experiments and Monte Carlo method. Using the calculation results of absorbed dose, the fast neutron inactivation effects on Bacillus subtilis were studied. The survival rates and absorbed dose curve was constructed. (authors)

  10. Neutron absorbing room temperature vulcanizable silicone rubber compositions

    International Nuclear Information System (INIS)

    Zoch, H.L.

    1979-01-01

    A neutron absorbing composition is described and consists of a one-component room temperature vulcanizable silicone rubber composition or a two-component room temperature vulcanizable silicone rubber composition in which the composition contains from 25 to 300 parts by weight based on the base silanol or vinyl containing diorganopolysiloxane polymer of a boron compound or boron powder as the neutron absorbing ingredient. An especially useful boron compound in this application is boron carbide. 20 claims

  11. Neutron absorbing article and method for manufacture of such article

    International Nuclear Information System (INIS)

    Hortman, M.T.; Mcmurtry, C.H.; Naum, R.G.; Owens, D.P.

    1980-01-01

    A neutron absorbing article, preferably in long, thin, flat form , suitable for but not necessarily limited to use in storage racks for spent nuclear fuel at locations between volumes of such stored fuel, to absorb neutrons from said spent fuel and prevent uncontrolled nuclear reaction of the spent fuel material, is composed of finely divided boron carbide particles and a solid, irreversibly cured phenolic polymer, forming a continuous matrix about the boron carbide particles, in such proportions that at least 6% of b10 from the boron carbide content is present therein. The described articles withstand thermal cycling from repeated spent fuel insertions and removals, withstand radiation from said spent nuclear fuel over long periods of time without losing desirable neutron absorbing and physical properties, are sufficiently chemically inert to water so as to retain neutron absorbing properties if brought into contact with it, are not galvanically corrodible and are sufficiently flexible so as to withstand operational basis earthquake and safe shutdown earthquake seismic events, without loss of neutron absorbing capability and other desirable properties, when installed in storage racks for spent nuclear fuel. The disclosure also relates to a plurality of such neutron absorbing articles in a storage rack for spent nuclear fuel and to a method for the manufacture of the articles

  12. A Study on the Design of Novel Neutron Absorber Using Artificial Rare Earth Compound

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Song Hyun; Shin, Chang Ho; Lee, Seung Hyun; Park, Jeia; Kim, Jong Kyung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Soon Young [RADCORE Co., Ltd., Daejeon (Korea, Republic of); Park, Hwan Seo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The artificial rare earth compounds (RE{sub 2}O{sub 3}) generated by the result of the pyro-processing are radioactive wastes which have many long-live radionuclides. Due to the high and long-lived radioactivity of the article RE{sub 2}O{sub 3}, specific radiation shielding and disposal techniques are required. In this study, a simultaneous disposal method of the RE{sub 2}O{sub 3} with the spent fuels is proposed by reusing them for the neutron absorber. In this study, the neutron absorber based on artificial RE{sub 2}O{sub 3} compound was designed for the use in the spent fuel storage. The design of the storage racks for the WH 17Χ17 and PLUS7 spent fuel assemblies were designed and the criticalities were evaluated with the various RE{sub 2}O{sub 3} compositions. Also, the radioactivity and irradiation calculations were performed for the applicability and stability analyses of the neutron absorber into the spent fuel storage. The results show that the neutron absorber can sufficiently reduce the criticality under the regulation guideline. It is expected that the neutron absorber can contribute minimizing the disposal area of the radioactive wastes as well as the reducing the costs and resources for the using the other types of the neutron absorbers.

  13. Neutron detector using sol-gel absorber

    Science.gov (United States)

    Hiller, John M.; Wallace, Steven A.; Dai, Sheng

    1999-01-01

    An neutron detector composed of fissionable material having ions of lithium, uranium, thorium, plutonium, or neptunium, contained within a glass film fabricated using a sol-gel method combined with a particle detector is disclosed. When the glass film is bombarded with neutrons, the fissionable material emits fission particles and electrons. Prompt emitting activated elements yielding a high energy electron contained within a sol-gel glass film in combination with a particle detector is also disclosed. The emissions resulting from neutron bombardment can then be detected using standard UV and particle detection methods well known in the art, such as microchannel plates, channeltrons, and silicon avalanche photodiodes.

  14. A neutron-absorbing porcelain enamel for coating nuclear equipment

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1988-01-01

    In 1985, nuclear safety analyses showed that under upset conditions, strict administrative controls were necessary to limit access to a new processing vessel for enriched uranium service at the Savannah River Plant (SRP). In order to increase the level of nuclear safety associated with that vessel, the traditional methods of incorporating neutron absorbers (borated stainless steel, boral, cadmium foil, etc.) were reviewed, however, process conditions did not permit their use. A neutron-absorbing porcelain enamel containing large amounts of cadmium and boron was developed as a safe, cost-effective alternative to traditional neutron-absorbing methods. Several pieces of coated process equipment have been installed or are planned for installation at SRP

  15. Miscellaneous neutron techniques

    International Nuclear Information System (INIS)

    Iddings, F.A.

    1976-01-01

    Attention is brought to the less often uses of neutrons in the areas of neutron radiography, well logging, and neutron gaging. Emphasis on neutron radiography points toward the isotopic sensitivity of the method versus the classical bulk applications. Also recognized is the ability of neutron radiography to produce image changes that correspond to thickness and density changes obtained in photon radiography. Similarly, neutron gaging applications center on the measurement of radiography. Similarly, neutron gaging applications center on the measurement of water, oil, or plastics in industrial samples. Well logging extends the neutron gaging to encompass many neutron properties and reactions besides thermalization and capture. Neutron gaging also gives information on organic structure and concentrations of a variety of elements or specific compounds in selected matrices

  16. Crystallographic structures of absorbates and neutron diffraction

    International Nuclear Information System (INIS)

    Marti, C.; Thorel, P.

    1975-01-01

    The advantage of neutron diffraction is that it is possible to work at any pressure and therefore to study an adsorbant-adsorbate couple within a wide pressure and temperature range and at thermodynamic equilibrium. Nitrogen adsorbed on graphite and CF 4 adsorbed on graphite were measured [fr

  17. Integrity of neutron-absorbing components of LWR fuel systems

    International Nuclear Information System (INIS)

    Bailey, W.J.; Berting, F.M.

    1991-03-01

    A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs

  18. Neutron radiography, techniques and applications

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1987-10-01

    After describing the principles of the ''in pool'' and ''dry'' installations, techniques used in neutron radiography are reviewed. Use of converter foils with silver halide films for the direct and transfer methods is described. Advantages of the use of nitrocellulose film for radiographying radioactive objects are discussed. Dynamic imaging is shortly reviewed. Standardization in the field of neutron radiography (ASTM and Euratom Neutron Radiography Working Group) is described. The paper reviews main fields of use of neutron radiography. Possibilities of use of neutron radiography at research reactors in various scientific, industrial and other fields are mentioned. Examples are given of application of neutron radiography in industry and the nuclear field. (author)

  19. Safety implications of anomalous effects of neutron absorbers on criticality

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1987-04-01

    A number of ''anomalies'' in nuclear criticality have been disclosed in recent years, and as new data have become available additional anomalies have come to light. Application of existing data, without familiarity with the anomalies could lead to diminished criticality control, or more costly less efficient control. As neutron absobers are frequently used for criticality control, this paper briefly presents and discusses six apparent anomalies pertaining to the effect of neutron absorbers on the criticality of fissionable material

  20. New techniques in neutron scattering

    International Nuclear Information System (INIS)

    Hayter, J.B.

    1993-01-01

    New neutron sources being planned, such as the Advanced Neutron Source (ANS) or the European Spallation Source (ESS), will provide an order of magnitude flux increase over what is available today, but neutron scattering will still remain a signal-limited technique. At the same time, the development of new materials, such as polymer and ceramic composites or a variety of complex fluids, will increasingly require neutron-based research. This paper will discuss some of the new techniques which will allow us to make better use of the available neutrons, either through improved instrumentation or through sample manipulation. Discussion will center primarily on unpolarized neutron techniques since polarized neutrons will be the subject of the next paper. (author)

  1. Neutron physics calculation for WWER-1000 absorber element lifetime determination

    International Nuclear Information System (INIS)

    Kurakin, K.Yu.; Kushmanov, S.A.

    2009-01-01

    Absorber element with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 - 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development. The paper deals with the issues related to calculation determination of main factors that influence AE service life limitation - neutron flux and fluence onto absorbing and structural materials during extended service life. (Authors)

  2. Pulse neutron logging technique

    International Nuclear Information System (INIS)

    Bespalov, D.F.; Dylyuk, A.A.

    1975-01-01

    A new method of neutron-burst logging is proposed, residing in irradiating rocks with fast neutron bursts and registering the integrated flux burst of thermal and/or epithermal neutrons, from the moment of its initiation to that of full absorption. The obtaained value is representative of the rock properties (porosity, hydrogen content). The integrated flux in a burst of thermal and epithermal neutrons can be measured both by way of activation of a reference sample of a known chemical composition during the neutron burst and by recording the radiation of induced activity of the sample within an interval between two bursts. The proposed method features high informative value, accuracy and efficiency

  3. A transformation technique to treat strong vibrating absorbers

    International Nuclear Information System (INIS)

    Sahni, D.C.; Garis, N.S.; Pazsit, I.

    1998-06-01

    Calculation of the neutron noise, induced by small amplitude vibrations of a strong absorber, is a difficult task because the traditional linearization technique cannot be applied. Two methods, based on two different representations of the absorber, were developed earlier to solve the problem. In both methods the rod displacements are described by a Taylor expansion, such that the boundary condition needs only to be considered at the surface of a static rod. Only one of the methods is applicable in two dimensions. In this paper an alternative method is developed and used for the solution of the problem. The essence of the method is a variable transformation by which the moving boundary is transformed into a static one without Taylor expansion. The corresponding equations are solved in a linear manner and the solution is transformed back to the original parameter space. The method is equally applicable in one and two dimensions. The solutions are in complete agreement with those of the previous methods

  4. Fluorescent converter and neutron absorber being made of boron nitride

    International Nuclear Information System (INIS)

    Matsumoto, G.; Teramura, M.; Sato, J.; Maeda, M.

    1983-01-01

    To improve the sensitivity of fluorescent converter is essential to the neutron radiography (NRG) which utilizes portable, not so strong, neutron sources. The fluorescent converter made of boron nitride (BN) is fabricated and tested. The sensitivity is about 1/20 of the NE426, but the homogeneity may be better. If 10 BN is utilized, the sensitivity will be five times as much as that of natural BN. Using the neutron beam of the Kyoto University Research Reactor, the flux of which is about 10 6 n/cm 2 sec, a good neutron television image was gained by X-ray television camera. As a bi-product of this converter, a flexible absorber was fabricated. (Auth.)

  5. Fuelling study of CANDU reactors using neutron absorber poisoned fuel

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.J.; Chan, P.K.; Bonin, H.W., E-mail: s25815@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2014-07-01

    A comparative fuelling study is conducted to determine the potential gain in operating margin for CANDU reactors incurred by implementing a change to the design of the conventional 37-element natural uranium (NU) fuel. The change involves insertion of minute quantities of neutron absorbers, Gd{sub 2}O{sub 3} and Eu{sub 2}O{sub 3}, into the fuel pellets. The Reactor Fuelling Simulation Program (RFSP) is used to conduct core-following simulations, for the regular 37-element NU fuel, which is to be used as control for comparison. Preliminary results are presented for fuelling with the regular 37-element NU fuel, which indicate constraints on fuelling that may be relaxed with addition of neutron absorbers. (author)

  6. Neutron absorber qualification and acceptance testing from the designer's perspective

    International Nuclear Information System (INIS)

    Bracey, W.; Chiocca, R.

    2004-01-01

    Starting in the mid 1990's, the USNRC began to require less than 100% credit for the 10B present in fixed neutron absorbers spent fuel transport packages. The current practice in the US is to use only 75% of the specified 10B in criticality safety calculations unless extensive acceptance testing demonstrates both the presence of the 10B and uniformity of its distribution. In practice, the NRC has accepted no more than 90% credit for 10B in recent years, while other national competent authorities continue to accept 100%. More recently, with the introduction of new neutron absorber materials, particularly aluminum / boron carbide metal matrix composites, the NRC has also expressed expectations for qualification testing, based in large part on Transnuclear's successful application to use a new composite material in the TN-68 storage / transport cask. The difficulty is that adding more boron than is really necessary to a metal has some negative effects on the material, reducing the ductility and the thermal conductivity, and increasing the cost. Excessive testing requirements can have the undesired effect of keeping superior materials out of spent fuel package designs, without a corresponding justification based on public safety. In European countries and especially in France, 100% credit has been accepted up to now with materials controls specified in the Safety Analysis Report (SAR): Manufacturing process approved by qualification testing Materials manufacturing controlled under a Quality Assurance system. During fabrication, acceptance testing directly on products or on representative samples. Acceptance criteria taking into account a statistical uncertainty corresponding to 3σ. The original and current bases for the reduced 10 B credit, the design requirements for neutron absorber materials, and the experience of Transnuclear and Cogema Logistics with neutron absorber testing are examined. Guidelines for qualification and acceptance testing and process controls

  7. Neutron absorbing article and method for manufacture thereof

    International Nuclear Information System (INIS)

    Forsyth, P.F.; Mcmurtry, C.H.; Naum, R.G.

    1980-01-01

    A composite, neutron absorbing, coated article, suitable for installation in storage racks for spent nuclear fuel and for other neutron absorbing applications, includes a backing member, preferably of flexible material such as woven fiberglass cloth, a synthetic organic polymeric coating or a plurality of such coatings on the backing member, preferably of cured phenolic resin, such as phenol formaldehyde or trimethylolphenol formaldehyde and boron carbide particles held to the backing member by the cured coating or a plurality of such coatings. Also within the invention is a method for the manufacture of the neutron absorbing coated article and the use of such an article. In a preferred method the backing member is first coated on both sides thereof with a filling coating of thermosettable liquid phenolic resin, which is then partially cured to solid state, one side of the backing member is then coated with a mixture of thermosettable liquid resin and finely divided boron carbide particles and the resin is partially cured to solid state, the other side is coated with a similar mixture, larger boron carbide particles are applied to it and the resin is partially cured to solid state, such side of the article is coated with thermosettable liquid phenolic resin, the resin is partially cured to solid state and such resin, including previously applied partially cured resins, is cured to final cross-linked and permanently set form

  8. A new neutron absorber material for criticality control

    International Nuclear Information System (INIS)

    Wells, Alan H.

    2007-01-01

    A new neutron absorber material based on a nickel metal matrix composite has been developed for applications such as the Transport, Aging, and Disposal (TAD) canister for the Yucca Mountain Project. This new material offers superior corrosion resistance to withstand the more demanding geochemical environments found in a 300,000 year to a million year repository. The lifetime of the TAD canister is currently limited to 10,000 years, reflecting the focus of current regulations embodied in 10 CFR 63. The use of DOE-owned nickel stocks from decommissioned enrichment facilities could reduce the cost compared to stainless steel/boron alloy. The metal matrix composite allows the inclusion of more than one neutron absorber compound, so that the exact composition may be adjusted as needed. The new neutron absorber material may also be used for supplementary criticality control of stored or transported PWR spent fuel by forming it into cylindrical pellets that can be inserted into a surrogate control rod. (authors)

  9. Experimental technique of neutron reflection

    International Nuclear Information System (INIS)

    Chen Bo; Huang Chaoqiang; Li Xinxi

    2006-12-01

    It is presented that the classifications, structures and components of neutron reflectometer (NR), as well s functions and parameters of each components, detailed characters of NR facility 'PRN-2M'. Based on the practical experiments, the basic experimental techniques, the measurement and the related experimental settings are described, including the choice of experimental conditions, adjustments of polarized neutron beam line, basic experimental technique and approach of measurement. The above can be an instruction for NR experiments and a reference for NR construction. (authors)

  10. Study of neutron absorbing microspheres in research reactors - Neutronic analyse

    International Nuclear Information System (INIS)

    Gana Watkins, Ignacio A.; Prado, Miguel O.; Mazufri, Claudio; Tunon, Juan M

    2012-01-01

    Now-a-days, it is increasingly common for nuclear power plants, as well as research reactors, to be designed and built with an alternative safety system aside from control rods. The acids and/or salts in solution injection systems is most frequently used. However, these systems present several implementation and operation problems due to the physical and chemical properties of the used compounds. After analyzing these drawbacks, we developed a new alternative safety system that contains the absorbing element isolated from the aqueous medium. In this context, it's proposed the use of aluminum borosilicate microspheres. The current paper presents erosion wear experiments to determine under which conditions microspheres can be considered as a potential component of a secondary shut down system in a nuclear facility (author)

  11. Characterization of weak, fair and strong neutron absorbing materials by means of neutron transmission: Beam hardening effect

    Science.gov (United States)

    Kharfi, F.; Bastuerk, M.; Boucenna, A.

    2006-09-01

    The characterization of neutron absorbing materials as well as quantification of neutron attenuation through matter is very essential in various fields, namely in shielding calculation. The objective of this work is to describe an experimental procedure to be used for the determination of neutron transmission through different materials. The proposed method is based on the relation between the gray value measured on neutron radiography image and the corresponding inducing neutron beam. For such a purpose, three kinds of materials (in shape of plate) were investigated using thermal neutrons: (1) boron-alloyed stainless steel as strong absorber; (2) copper and steel as fair absorbers and (3) aluminum as weak absorber. This work is not limited to the determination of neutron transmission through matters; it is also spread out to the measure of the surface density of the neutron absorbing elements (ρs) as a function of thickness of neutron absorbing material such as boron-alloyed stainless steel. The beam hardening effect depending on material thickness was also studied using the neutron transmission measurements. A theoretical approach was used to interpret the experimental results. The neutron transmission measurements were performed at the Neutron Radiography and Tomography facility of the Atomic Institute of the Austrian Universities in Vienna. Finally, a Maxwellian neutron distribution of incident neutron beam was used in the theoretical calculations of neutron energy shift in order to compare with experiments results. The obtained experimental results are in a good agreement with the developed theoretical approach.

  12. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1982-11-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone TE-solutions, mineral oil and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. the OAR's measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. It is recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. A table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  13. Neutron absorbing article and method for manufacture of such article

    International Nuclear Information System (INIS)

    McMurty, C.H.; Naum, R.G.; Owens, D.P.; Hortman, M.T.

    1981-01-01

    A neutron absorbing article is described which comprises boron carbide particles and an irreversibly-cured phenol aldehyde condensation polymer cured to a continuous matrix about the boron carbide particles. Such an article may be used in spent fuel storage racks. It can be manufactured by mixing together a curable phenolic resin with boron carbide particles, compacting the mixture to an article of desired shape, curing the resin at an elevated temperature, impregnating the cured article with curable phenolic resin in liquid state, and curing the article again

  14. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1983-01-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone tissue-equivalent (TE) solutions, mineral oil, and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU Report No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. The OARs measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. Therefore, neutron beam CADDs and OARs may be measured in either TE solution (USA practice) or water (European practice), and having determined the respective scaling lengths, all measurements may be scaled from one medium to any other. It is recommended that for general treatment planning purposes, scaling be made to TE muscle with a density of 1.04 g cm -3 , since this value represents muscle and other soft tissues better than TE solution of density 1.07 g cm -3 . For such a transformation, relative measurements made in water are found to require very small corrections. Hence, it is further recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. Finally, a table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  15. New Techniques in Neutron Scattering

    DEFF Research Database (Denmark)

    Birk, Jonas Okkels

    potential performance than any existing facility, however in order to use this pulse structure optimally many existing neutron scattering instruments will need to be redesigned. This defense will concentrate on the design and optimization of the inverse time-of-flight cold neutron spectrometer CAMEA......, simulations and prototyping to optimize the instrument and ensure that it will deliver the predicted performance when constructed. During the design a new prismatic analyser concept that can be of interest to many other neutron spectrometers was developed. The design work was compiled into an instrument......Neutron scattering is an important experimental technique in amongst others solid state physics, biophysics, and engineering. This year construction of European Spallation Source (ESS) was commenced in Lund, Sweeden. The facility will use a new long pulsed source principle to obtain higher...

  16. Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

    International Nuclear Information System (INIS)

    Rossa, Riccardo; Borella, Alessandro; Labeau, Pierre-Etienne; Pauly, Nicolas; Meer, Klaas van der

    2015-01-01

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239 Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239 Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239 Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235 U fission chamber, with a 3 He proportional counter, and with a 10 B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239 Pu and 235 U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3 He and 10 B proportional counters to increase the sensitivity to 239 Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies

  17. Apparatus and method for the measurement of neutron moderating or absorbing properties of objects

    International Nuclear Information System (INIS)

    Untermyer, S.I.

    1981-01-01

    An apparatus and method for measuring the neutron moderating or absorbing properties of objects or materials is disclosed in which a fast neutron source cooperates with a neutron absorbing material which reduces the energy of the fast neutrons by inelastic scattering so that they can be readily thermalized by a moderator. A thermal neutron detector is disposed adjacent the material and serves to detect thermal neutrons emitted by a moderator placed to receive and thermalize the reduced energy neutrons. A material whose absorption is to be measured is placed between a moderator and the detector

  18. Scram device having a multiplicity of neutron absorbing masses

    International Nuclear Information System (INIS)

    Giuggio, N.; Noyes, R.C.

    1981-01-01

    An apparatus is described for holding, releasing, and resetting a multiplicity of neutron-absorbing balls within a safety assembly of a liquid metal reactor. Vertically-hinged trap doors rest on the shoulders of a generally cylindrical release valve which is actuated by either the regular or by the self-actuated scram actuator. The doors and the valve shoulder provide a floor for the balls to be suspended above the reactor core during normal operation. When the actuator displaces the release valve, the doors lose their support and swing downward, permitting the poison balls to drop into the core. In the reset mode of operation, a platform at the bottom of the core is raised to lift the balls and swing the trap doors upward until the balls are above the door hinges. The release valve is reset to support the doors and the platform is lowered to the bottom of the safety assembly

  19. Performance evaluation of METAMIC neutron absorber in spent fuel storage rack

    Directory of Open Access Journals (Sweden)

    Kiyoung Kim

    2018-06-01

    Full Text Available High-density spent fuel (SF storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others. Keywords: Blister, Criticality, METAMIC, Neutron Absorber, Neutron Attenuation Test, Scanning Electron Microscope

  20. Pulsed neutron generator for use with pulsed neutron activation techniques

    International Nuclear Information System (INIS)

    Rochau, G.E.

    1980-01-01

    A high-output, transportable, pulsed neutron generator has been developed by Sandia National Laboratories for use with Pulsed Neutron Activation (PNA) techniques. The PNA neutron generator generates > 10 10 14 MeV D-T neutrons in a 1.2 millisecond pulse. Each operation of the unit will produce a nominal total neutron output of 1.2 x 10 10 neutrons. The generator has been designed to be easily repaired and modified. The unit requires no additional equipment for operation or measurement of output

  1. Nuclear reactor control device by vertical displacement of neutron absorber scram rods

    International Nuclear Information System (INIS)

    Defaucheux, Jacques; Pasqualini, Gilbert; Wiart, Albert; Martin, Jean.

    1981-01-01

    Nuclear reactor control system by vertical displacement of an assembly absorbing the neutrons inside a reactor core and drop of the absorbing assembly in maximum insertion position under the effect of its own weight for emergency shutdown. The absorbing assembly is secured to the bottom end of a vertical control rod, the displacement of which is actuated by an electro-magnetic device [fr

  2. Genetic effects induced by neutrons in Drosophila melanogaster I. Determination of absorbed dose

    International Nuclear Information System (INIS)

    Delfin, A.; Paredes, L.C.; Zambrano, F.; Guzman-Rincon, J.; Urena-Nunez, F.

    2001-01-01

    A method to obtain the absorbed dose in Drosophila melanogaster irradiated in the thermal column facility of the Triga Mark III Reactor has been developed. The method is based on the measurements of neutron activation of gold foils produced by neutron capture to obtain the neutron fluxes. These fluxes, combined with the calculations of kinetic energy released per unit mass, enables one to obtain the absorbed doses in Drosophila melanogaster

  3. Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    Rossa, Riccardo, E-mail: rrossa@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Borella, Alessandro, E-mail: aborella@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Labeau, Pierre-Etienne, E-mail: pelabeau@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Pauly, Nicolas, E-mail: nipauly@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Meer, Klaas van der, E-mail: kvdmeer@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium)

    2015-08-11

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of {sup 239}Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a {sup 239}Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to {sup 239}Pu, in comparison with a {sup 235}U fission chamber, with a {sup 3}He proportional counter, and with a {sup 10}B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the {sup 239}Pu and {sup 235}U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the {sup 3}He and {sup 10}B proportional counters to increase the sensitivity to {sup 239}Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies.

  4. Modern techniques of structural neutron diffraction

    International Nuclear Information System (INIS)

    Aksenov, V.L.; )

    1997-01-01

    Modern techniques of neutron diffraction for structural investigations are analyzed. The time-of-flight method and the reverse time-of-flight method are considered briefly. Characteristics of two-crystal and time-of-flight neutron diffractometers are compared. It is pointed that in the future, the great importance will be possessed the development of high-resolution Fourier neutron diffractometers [ru

  5. Thermal Evaluation of Storage Rack with an Advanced Neutron Absorber during Normal Operation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Jae; Kim, Mi-Jin; Sohn, Dong-Seong [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The storage capacity of the domestic wet storage site is expected to reach saturation from Hanbit in 2024 to Sin-wolseong in 2038 and accordingly management alternatives are urgently taken. Since installation of the dense rack is considered in the short term, it is necessary to urgently develop an advanced neutron absorber which can be applied to a spent nuclear fuel storage facility. Neutron absorber is the material for controlling the reactivity. A material which has excellent thermal neutron absorption ability, high strength and corrosion resistance must be selected as the neutron absorber. Existing neutron absorbers are made of boron which has a good thermal absorption ability such as BORAL and METAMIC. However, possible problems have been reported in using the boron-based neutron absorber for wet storage facility. Gadolinium is known to have higher neutron absorption cross-section than that of boron. And the strength of duplex stainless steel is about 1.5 times higher than stainless steel 304 which has been frequently used as a structural material. Therefore, duplex stainless steel which contains gadolinium is in consideration as an advanced neutron absorber. Temperature distribution is shown in figure 4. In pool bottom region near the inlet shows a relatively low tendency and heat generated from the fuel assemblies is transmitted to the pool upper region by the vertical flow. Also, temperature gradient appear in rack structures for the axial direction and temperature is uniformly distributed in the pool upper region. Table 1 presents the calculated results. The maximum temperature is 306.63K and does not exceed the 333.15K (60℃). The maximum temperature of the neutron absorber is 306.48K.

  6. Study of thermal neutron currents near cylindrical absorbers located in heavy water

    International Nuclear Information System (INIS)

    Simard, Y.N.

    1973-01-01

    The experiments reported involved determining the angular response of detectors to neutrons exterior to the surface of long cylindrical absorbers immersed in a scattering medium. The absorbers consisted of solid cylinders of copper, cadmium, or natural uranium in a fuel lattice, and combinations of copper and cadmium, as well as voided cylinders. The scattering (moderating) medium consisted of heavy water. (author)

  7. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

    1988-11-01

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  8. Neutron relative biological effectiveness for solid cancer incidence in the Japanese A-bomb survivors: an analysis considering the degree of independent effects from γ-ray and neutron absorbed doses with hierarchical partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Linda [Federal Office for Radiation Protection, Department Radiation Protection and Health, Oberschleissheim (Germany); University of Manchester, The Faculty of Medical and Human Sciences, Manchester (United Kingdom)

    2013-03-15

    It has generally been assumed that the neutron and γ-ray absorbed doses in the data from the life span study (LSS) of the Japanese A-bomb survivors are too highly correlated for an independent separation of the all solid cancer risks due to neutrons and due to γ-rays. However, with the release of the most recent data for all solid cancer incidence and the increased statistical power over previous datasets, it is instructive to consider alternatives to the usual approaches. Simple excess relative risk (ERR) models for radiation-induced solid cancer incidence fitted to the LSS epidemiological data have been applied with neutron and γ-ray absorbed doses as separate explanatory covariables. A simple evaluation of the degree of independent effects from γ-ray and neutron absorbed doses on the all solid cancer risk with the hierarchical partitioning (HP) technique is presented here. The degree of multi-collinearity between the γ-ray and neutron absorbed doses has also been considered. The results show that, whereas the partial correlation between the neutron and γ-ray colon absorbed doses may be considered to be high at 0.74, this value is just below the level beyond which remedial action, such as adding the doses together, is usually recommended. The resulting variance inflation factor is 2.2. Applying HP indicates that just under half of the drop in deviance resulting from adding the γ-ray and neutron absorbed doses to the baseline risk model comes from the joint effects of the neutrons and γ-rays - leaving a substantial proportion of this deviance drop accounted for by individual effects of the neutrons and γ-rays. The average ERR/Gy γ-ray absorbed dose and the ERR/Gy neutron absorbed dose that have been obtained here directly for the first time, agree well with previous indirect estimates. The average relative biological effectiveness (RBE) of neutrons relative to γ-rays, calculated directly from fit parameters to the all solid cancer ERR model with both

  9. Mitigation of end flux peaking in CANDU fuel bundles using neutron absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, D.; Chan, P.K., E-mail: dylan.pierce@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Shen, W. [Canadian Nuclear Safety Commission, Ottawa ON, (Canada)

    2015-07-01

    End flux peaking (EFP) is a phenomenon where a region of elevated neutron flux occurs between two adjoining fuel bundles. These peaks lead to an increase in fission rate and therefore greater heat generation. It is known that addition of neutron absorbers into fuel bundles can help mitigate EFP, yet implementation in Canada Deuterium Uranium (CANDU) type reactors using natural uranium fuel has not been pursued. Monte Carlo N-Particle code (MCNP) 6.1 was used to simulate the addition of a small amount of neutron absorbers strategically within the fuel pellets. This paper will present some preliminary results collected thus far. (author)

  10. Processing requirements for property optimization of Eu2O3-W cermets for fast reactor neutron absorber applications

    International Nuclear Information System (INIS)

    Pasto, A.E.; Tennery, V.J.

    1977-01-01

    Europium sesquioxide is a candidate fast reactor neutron absorber material. It possesses several desirable characteristics for this application, but has a low thermal conductivity. This gives rise to pellet cracking during reactor operation. To increase the thermal conductivity without great sacrifice in nuclear worth, addition of tungsten to Eu 2 O 3 has been evaluated. Synthesis and fabrication techniques described allow preparation of high density compacts of Eu 2 O 3 -15 vol. percent tungsten, possessing favorable thermal conductivity and thermal expansion characteristics

  11. Long-term effects of neutron absorber and fuel matrix corrosion on criticality

    International Nuclear Information System (INIS)

    Culbreth, W.G.; Zielinski, P.R.

    1994-01-01

    Proposed waste package designs will require the addition of neutron absorbing material to prevent the possibility of a sustained chain reaction occurring in the fuel in the event of water intrusion. Due to the low corrosion rates of the fuel matrix and the Zircaloy cladding, there is a possibility that the neutron absorbing material will corrode and leak from the waste container long before the subsequent release of fuel matrix material. An analysis of the release of fuel matrix and neutron absorber material based on a probabilistic model was conducted and the results were used to prepare input to KENO-V, an neutron criticality code. The results demonstrate that, in the presence of water, the computed values of k eff exceeded the maximum of 0.95 for an extended period of time

  12. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  13. Process and device for exchanging neutron absorber rods

    International Nuclear Information System (INIS)

    Baero, G.; Kraus, W.; Stindt, W.

    1987-01-01

    The control element repair device contains lifting equipment for inserting the control element in the accommodation device. Due to the case position assigned to each absorber rod of a control element, after removing the carrier with the absorber rods fixed to it, the defective rods can be replaced by new ones. The accommodation device has a support to support the carrier. Turning the control element for the PWR through 180 0 is prevented. (DG) [de

  14. Neutron flux measurement utilizing Campbell technique

    International Nuclear Information System (INIS)

    Kropik, M.

    2000-01-01

    Application of the Campbell technique for the neutron flux measurement is described in the contribution. This technique utilizes the AC component (noise) of a neutron chamber signal rather than a usually used DC component. The Campbell theorem, originally discovered to describe noise behaviour of valves, explains that the root mean square of the AC component of the chamber signal is proportional to the neutron flux (reactor power). The quadratic dependence of the reactor power on the root mean square value usually permits to accomplish the whole current power range of the neutron flux measurement by only one channel. Further advantage of the Campbell technique is that large pulses of the response to neutrons are favoured over small pulses of the response to gamma rays in the ratio of their mean square charge transfer and thus, the Campbell technique provides an excellent gamma rays discrimination in the current operational range of a neutron chamber. The neutron flux measurement channel using state of the art components was designed and put into operation. Its linearity, accuracy, dynamic range, time response and gamma discrimination were tested on the VR-1 nuclear reactor in Prague, and behaviour under high neutron flux (accident conditions) was tested on the TRIGA nuclear reactor in Vienna. (author)

  15. Role of cytogenetic techniques in biological dosimetry of absorbed radiation

    International Nuclear Information System (INIS)

    Rao, B.S.

    2016-01-01

    In most of the radiation accidents, physical dosimetric information is rarely available. Further, most of the accidental exposures are non-uniform involving either partial body or localized exposure to significant doses. In such situations, physical dosimetry does not provide reliable dose estimate. It has now been realized that biological dosimetric techniques can play an important role in the assessment of absorbed dose. In recent years, a number of biological indicators of radiation have been identified. These include the kinetics of onset and persistence of prodromal syndromes (radiation sickness), cytogenetic changes in peripheral blood lymphocytes, hematological changes, biochemical indicators, ESR spectroscopy of biological samples, induction of gene mutations in red blood cells, cytogenetic and physiological changes in skin and neurophysiological changes. In general, dosimetric information is derived by a combination of several different methods, as they have potential to serve as prognostic indicators. The role of cytogenetic techniques in peripheral blood lymphocytes (PBL) as biological indicators of absorbed radiation is reviewed here

  16. Absorbed dose conversion coefficients for embryo and foetus in neutron fields

    International Nuclear Information System (INIS)

    Chen, J.

    2007-01-01

    The Monte Carlo code MCNPX has been used to determine mean absorbed doses to the embryo and foetus when the mother is exposed to neutron fields. There are situations, such as on-board aircraft, where high-energy neutrons are often peaked in top down (TOP) direction. In addition to previous publications for standard irradiation geometries, this study provides absorbed dose conversion coefficients for the embryo of 8 weeks and the foetus of 3, 6 or 9 months at TOP irradiation geometry. The conversion coefficients are compared with the coefficients in isotropic irradiation (ISO). With increasing neutron energies, the conversion coefficients in TOP irradiation become dominant. A set of conversion coefficients is constructed from the higher value in either ISO or TOP irradiation at a given neutron energy. In cases where the irradiation geometry is not adequately known, this set of conversion coefficients can be used in a conservative dose assessment for embryo and foetus in neutron fields. (authors)

  17. Excitation of surface waves of ultracold neutrons on absorbing trap walls as anomalous loss factor

    International Nuclear Information System (INIS)

    Bokun, R.Ch.

    2006-01-01

    One analyzed probability of excitation of surface waves of ultracold neutrons in terms of a plane model consisting of three media: vacuum, a finite depth neutron absorbing substance layer and a neutron reflecting substrate. One demonstrated the absence of the mentioned surface waves in terms of the generally accepted model of two media: vacuum contiguous to the plane surface of a substance filled half-space. One pointed out the effect of the excited surface waves of ultracold neutrons on the increase of their anomalous losses in traps [ru

  18. Neutron borehole logging correction technique

    International Nuclear Information System (INIS)

    Goldman, L.H.

    1978-01-01

    In accordance with an illustrative embodiment of the present invention, a method and apparatus is disclosed for logging earth formations traversed by a borehole in which an earth formation is irradiated with neutrons and gamma radiation produced thereby in the formation and in the borehole is detected. A sleeve or shield for capturing neutrons from the borehole and producing gamma radiation characteristic of that capture is provided to give an indication of the contribution of borehole capture events to the total detected gamma radiation. It is then possible to correct from those borehole effects the total detected gamma radiation and any earth formation parameters determined therefrom

  19. Pulsed neutron activation calibration technique

    International Nuclear Information System (INIS)

    Kehler, P.

    1979-01-01

    A pulsed neutron activation (PNA) for measurement of two-phase flow consists of a pulsed source of fast neutron to activate the oxygen in a steam-water mixture. Flow is measured downstream by an NaI detector. Measured counts are sorted by a multiscaler into different time channels. A counts vs. time distribution typical for two-phase flow with slip between the two phases is obtained. Proper evaluation for the counts/time distribution leads to flow-regime independent equations for the average of the inverse transil time and the average density. After calculation of the average mass flow velocity, the true mass flow is derived

  20. 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

  1. Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

    Energy Technology Data Exchange (ETDEWEB)

    Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

    2016-09-15

    Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

  2. 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

  3. Determination of hydrogen content by neutron techniques

    International Nuclear Information System (INIS)

    Santisteban, J.R.; Granada, J.R.; Mayer, R.E.

    1997-01-01

    The commonly available techniques for the determination of hydrogen dissolved in solids are usually destructive from the point of view of the sample. A new, nondestructive method for this kind of measurements has been developed at our laboratory, with the requirement of improved sensitivity for massive samples. This scattering method is based on the use of epithermal neutrons, and has been implemented through the design and construction of a spectrometer dedicated to that task. In addition, the traditional transmission method has been employed to determine hydrogen content in metals, using the full sub thermal and thermal neutron energy ranges. A pulsed neutron source based on an electron LINAC is employed, together with time-of-flight techniques. In this work we will present some results illustrative of the sensitivity achieved by these neutron techniques in different systems and for a wide range of hydrogen concentrations. (author) [es

  4. Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corcoran, E.C. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada)], E-mail: emily.corcoran@rmc.ca; Lewis, B.J.; Thompson, W.T. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada); Hood, J. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada); Akbari, F.; He, Z. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ont., K0J 1J0 (Canada); Reid, P. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada)

    2009-03-31

    Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU [CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

  5. 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.

  6. Neutron measurement techniques for tokamak plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1994-01-01

    The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)

  7. Techniques in high pressure neutron scattering

    CERN Document Server

    Klotz, Stefan

    2013-01-01

    Drawing on the author's practical work from the last 20 years, Techniques in High Pressure Neutron Scattering is one of the first books to gather recent methods that allow neutron scattering well beyond 10 GPa. The author shows how neutron scattering has to be adapted to the pressure range and type of measurement.Suitable for both newcomers and experienced high pressure scientists and engineers, the book describes various solutions spanning two to three orders of magnitude in pressure that have emerged in the past three decades. Many engineering concepts are illustrated through examples of rea

  8. Neutron Scattering in Biology Techniques and Applications

    CERN Document Server

    Fitter, Jörg; Katsaras, John

    2006-01-01

    The advent of new neutron facilities and the improvement of existing sources and instruments world wide supply the biological community with many new opportunities in the areas of structural biology and biological physics. The present volume offers a clear description of the various neutron-scattering techniques currently being used to answer biologically relevant questions. Their utility is illustrated through examples by some of the leading researchers in the field of neutron scattering. This volume will be a reference for researchers and a step-by-step guide for young scientists entering the field and the advanced graduate student.

  9. Investigation of reactivity change and neutron noise due to random absorber vibrations. 2

    International Nuclear Information System (INIS)

    Barthel, R.

    1984-01-01

    Perturbations of the neutron flux due to stochastically excited vibrations of absorbers have been investigated using a one-dimensional core model with N pointlike absorbers. Taking into account the flux depressions near the absorbers, pronounced peaks in the spectral power densities of the flux fluctuations have been found at multiples of the resonance frequencies in addition to the direct imaging of the resonances of absorber vibrations. Investigation of the space dependence of the corresponding transfer functions has shown that a localization is possible by means of the double frequency effect and that the dispersion of absorber vibrations can be determined by using the triple frequency effect. The conclusions of the paper are qualitatively compared with results of noise measurements at a pressurized water reactor. (author)

  10. Carbon filter property detection with thermal neutron technique

    International Nuclear Information System (INIS)

    Deng Zhongbo; Han Jun; Li Wenjie

    2003-01-01

    The paper discussed the mechanism that the antigas property of the carbon filter will decrease because of its carbon bed absorbing water from the air while the carbon filter is being stored, and introduced the principle and method of detection the amount of water absorption with thermal neutron technique. Because some certain relation between the antigas property of the carbon filter and the amount of water absorption exists, the decrease degree of the carbon filter antigas property can be estimated through the amount of water absorption, offering a practicable facility technical pathway to quickly non-destructively detect the carbon filter antigas property

  11. Process and device for identifying nuclear reactor neutron absorber rod etancheity defect

    International Nuclear Information System (INIS)

    Pelletier, J.; Parrat, D.

    1990-01-01

    For identifying defects in the sealing of neutron absorbing rods. The rod is placed in a pressure tight enclosure filled with a chemically agressive solution. After a time the pressure is released to allow the solution come out of the rod. An analysis of the solution allows the detection of radioactive isotopes of metals which are in the rod [fr

  12. Applications of neutron activation analysis technique

    International Nuclear Information System (INIS)

    Jonah, S. A.

    2000-07-01

    The technique was developed as far back as 1936 by G. Hevesy and H. Levy for the analysis of Dy using an isotopic source. Approximately 40 elements can be analyzed by instrumental neutron activation analysis (INNA) technique with neutrons from a nuclear reactor. By applying radiochemical separation, the number of elements that can be analysed may be increased to almost 70. Compared with other analytical methods used in environmental and industrial research, NAA has some unique features. These are multi-element capability, rapidity, reproducibility of results, complementarity to other methods, freedom from analytical blank and independency of chemical state of elements. There are several types of neutron sources namely: nuclear reactors, accelerator-based and radioisotope-based sources, but nuclear reactors with high fluxes of neutrons from the fission of 235 U give the most intense irradiation, and hence the highest available sensitivities for NAA. In this paper, the applications of NAA of socio-economic importance are discussed. The benefits of using NAA and related nuclear techniques for on-line applications in industrial process control are highlighted. A brief description of the NAA set-ups at CERT is enumerated. Finally, NAA is compared with other leading analytical techniques

  13. Neutron radiographic techniques, facilities and applications

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1984-08-01

    This is a collection of three papers, written for presentation on two international conferences. The first paper: ''Neutron radiography. Techniques and facilities'', written by J.P. Barton of N-Ray Engineering Co. La Jolla, CA., USA and J.C. Domanus was presented at the International Symposium on the Use and Development of Low and Medium Flux Research Reactors at the Massachusets Institute of Technology, Cambridge, Mass., USA, 16-19 October 1983. The second paper: ''Neutron radiography with the DR-1 reactor at Risoe National Laboratory'', written by J.C. Domanus, was presented at the same Symposium. The third paper: ''Defects in nuclear fuel revealed by neutron radiography'', written by J.C. Domanus is accepted for presentation on 18 October 1984 to the 3rd European Conference on Nondestructive Testing, Florence, Italy, 15-18 October 1984. While the first paper describes the principles of neutron radiographic techniques and facilities, the second one describes an example of such facility and the third gives an example of application of neutron radiography in the field of nuclear fuel. (author)

  14. Understanding and predicting the behaviour of silver base neutron absorbers under irradiations

    International Nuclear Information System (INIS)

    Desgranges, C.

    1998-01-01

    The effect of neutron irradiation induced transmutations on the swelling of AgInCd (AIC) alloys used as neutron absorber in the control rods of Pressurized Water Reactors has been studied both experimentally and theoretically. Effective atomic volumes have been determined in synthetic AgCdInSn alloys with various compositions and containing fcc and hc phases, representative of irradiated AIC (Sn is a transmutation product). Swelling is shown to result first from the transmutation of Ag into Cd and of In into Sn, both with larger effective volume than the mother atom, and second from grain boundaries precipitation of s still less dense hc phase when solid solubility of transmuted products is exceeded. For both fcc and hc phases, we have determined profiles at the temperatures in the vicinity of the operating temperature. Unusual characteristics of second phase growth at grain boundaries induced by transmutations are identified on a simple binary alloy model: kinetics is controlled by irradiation temperature which scales diffusivities and flux which scales transmutation rates, as well as by the grain size in the underlying matrix. To address the AgInCdSn alloys, a novel technique is proposed to model diffusion in multicomponent alloys. It is based on a linearization of a simple atomistic model. With a single set of parameters, for each phase, our model well reproduces our interdiffusion measurements in quaternary alloys as well as existing interdiffusion experiments in binary alloys. Finally this diffusion model implemented with a moving interface algorithm is used to model the growth of the second phase induced by transmutation in the AIC under irradiation. (authors)

  15. Integrating techniques for neutron dosimetry in Linac 18 MV

    International Nuclear Information System (INIS)

    Ceron R, P. V.; Diaz G, J. A. I.; Rivera M, T.; Paredes G, L. C.; Vega C, H. R.

    2015-10-01

    In this paper thermoluminescent dosimetry, analytical techniques and Monte Carlo calculations were used to estimate the neutron dose equivalent in a radiotherapy room with a linear electron accelerator of 18 MV. The equivalent dose was measured at isocenter to 1.42 m of target and at the entrance of the labyrinth of the room of a Novalis Tx. The neutron detectors were constructed with pairs of thermoluminescent dosimeters TLD 600 ( 6 LiF: Mg, Ti) and TLD 700 ( 7 LiF: Mg, Ti) which are placed inside a paraffin sphere of 20 cm in diameter. These measurements enabled the calculation of equivalent dose in the gate and the source term, using the relationships contained in the NCRP-151. Through the models carried out with the code MCNPX the absorbed dose distribution with regard to depth in a paraffin phantom are included and the neutron spectrum produced by the head, taking into account the geometry and component materials. The results are in the order of neutron milli sievert by gray of X-rays (mSv/Gy x) which are in the same order as those found in other reports for different accelerators. (Author)

  16. Simulation of a silicon neutron detector coated with TiB2 absorber

    International Nuclear Information System (INIS)

    Krapohl, D; Nilsson, H-E; Petersson, S; Slavicek, T; Thungström, G; Pospisil, S

    2012-01-01

    Neutron radiation cannot be directly detected in semiconductor detectors and therefore needs converter layers. Planar clean-room processing can be used in the manufacturing process of semiconductor detectors with metal layers to produce a cost-effective device. We used the Geant4 Monte-Carlo toolkit to simulate the performance of a semiconductor neutron detector. A silicon photo-diode was coated with vapour deposited titanium, aluminium thin films and a titaniumdiboride (TiB 2 ) neutron absorber layer. The neutron capture reaction 10B(n, alpha)7Li is taken advantage of to create charged particles that can be counted. Boron-10 has a natural abundance of about SI 19.8%. The emitted alpha particles are absorbed in the underlying silicon detector. We varied the thickness of the converter layer and ran the simulation with a thermal neutron source in order to find the best efficiency of the TiB 2 converter layer and optimize the clean room process.

  17. Some neutron absorbing elements and devices for fast nuclear reactors regulation systems

    International Nuclear Information System (INIS)

    Kervalishvili, P.J.

    2010-01-01

    It is shown that performed technological, physical-mechanical and radiation tests clearly indicate the prospects of using Neutron Absorbing Elements (NAE) based on B-10 and some rare-earth compounds during the creation of highly effective Control and Safety System (CSS) rods for fast neutron nuclear energetic reactors. Particular attention was paid to the development of new and upgrading of existing computing and real technologies for designing and preparing the optimizing NAE items characterized by all physical and strength properties for obtaining desirable operational parameters of CSS rods on their base

  18. Neutron absorber qualification and acceptance testing from the designer's perspective

    Energy Technology Data Exchange (ETDEWEB)

    Bracey, W. [Transnuclear, Inc, Hawthorne, NY (United States); Chiocca, R. [Cogema Logistics, St. Quentin en Yvelines (France)

    2004-07-01

    Starting in the mid 1990's, the USNRC began to require less than 100% credit for the 10B present in fixed neutron absorbers spent fuel transport packages. The current practice in the US is to use only 75% of the specified 10B in criticality safety calculations unless extensive acceptance testing demonstrates both the presence of the 10B and uniformity of its distribution. In practice, the NRC has accepted no more than 90% credit for 10B in recent years, while other national competent authorities continue to accept 100%. More recently, with the introduction of new neutron absorber materials, particularly aluminum / boron carbide metal matrix composites, the NRC has also expressed expectations for qualification testing, based in large part on Transnuclear's successful application to use a new composite material in the TN-68 storage / transport cask. The difficulty is that adding more boron than is really necessary to a metal has some negative effects on the material, reducing the ductility and the thermal conductivity, and increasing the cost. Excessive testing requirements can have the undesired effect of keeping superior materials out of spent fuel package designs, without a corresponding justification based on public safety. In European countries and especially in France, 100% credit has been accepted up to now with materials controls specified in the Safety Analysis Report (SAR): Manufacturing process approved by qualification testing Materials manufacturing controlled under a Quality Assurance system. During fabrication, acceptance testing directly on products or on representative samples. Acceptance criteria taking into account a statistical uncertainty corresponding to 3{sigma}. The original and current bases for the reduced {sup 10}B credit, the design requirements for neutron absorber materials, and the experience of Transnuclear and Cogema Logistics with neutron absorber testing are examined. Guidelines for qualification and acceptance testing and

  19. Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

  20. Apparatus for controlling a nuclear reactor by vertical displacement of a unit absorbing neutrons

    International Nuclear Information System (INIS)

    Wiart, A.; Defaucheux, J.; Martin, J.; Pasqualini, G.

    1980-01-01

    Apparatus is described for controlling a nuclear reactor by vertical displacement of a unit absorbing neutrons, comprising, inside a sealed enclosure in communication with the interior of the reactor, a movable magnetic piece connected to a control shaft which is itself connected to the absorbent unit. This magnetic piece has at least two radial projections. The magnetic piece is displaced by an inductor with at least two pole shoes corresponding to the projections on the magnetic piece and allowing magnetic coupling between the inductor and the magnetic piece. The inductor and its displacement device are disposed outside the sealed enclosure. A control means allows the control shaft to be uncoupled from a member assuring its suspension so as to drop the absorbent unit in the event of emergency shutdown. The apparatus is particularly applicable to control rods of pressurized water nuclear reactors

  1. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-01-01

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution's concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the 'Poisoned Tube Tank' because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service

  2. Neutron visual sensing techniques making good use of computer science

    International Nuclear Information System (INIS)

    Kureta, Masatoshi

    2009-01-01

    Neutron visual sensing technique is one of the nondestructive visualization and image-sensing techniques. In this article, some advanced neutron visual sensing techniques are introduced. The most up-to-date high-speed neutron radiography, neutron 3D CT, high-speed scanning neutron 3D/4D CT and multi-beam neutron 4D CT techniques are included with some fundamental application results. Oil flow in a car engine was visualized by high-speed neutron radiography technique to make clear the unknown phenomena. 4D visualization of pained sand in the sand glass was reported as the demonstration of the high-speed scanning neutron 4D CT technique. The purposes of the development of these techniques are to make clear the unknown phenomena and to measure the void fraction, velocity etc. with high-speed or 3D/4D for many industrial applications. (author)

  3. Neutron physical investigations on the use of burnable poisons and gray absorber rods in large pressurized water reactors

    International Nuclear Information System (INIS)

    Brosche, C.; Katinger, T.; Kollmar, W.; Thieme, K.; Wagner, M.R.

    1977-11-01

    Methods and results of neutron physics calculations are described using burnable poisons and gray absorber rods in large PWR's. Calculated and measured values are compared, the effort for programming has been guessed. (orig.) [de

  4. A technique of measuring neutron spectrum

    International Nuclear Information System (INIS)

    Sarkar, P.K.; Kirthi, K.N.; Ganguly, A.K.

    1975-01-01

    Plastic scintillators have been used to measure fast neutron spectrum from various sources. Gamma background discrimination has been done by selecting thin scintillators and thereby achieving near 100% transmission of Compton-edge electrons. The measured distribution has been unfolded by using an iterative least square technique. This gives minimum variance and maximum likelihood estimate with error minimised. Smoothening of the observed distribution has been done by Fourier and time series analyses. The method developed is applicable in principle for the determination of spectra of high energy neutrons ranging from 1 MeV to 70 MeV and beyond. However, practical application of the method is limited by the non-availability of cross-section data for various neutron induced reactions with carbon and hydrogen present in the polymerised polystyrene scintillator. This procedure has been adopted in the present work for spectral determination up to 14 MeV neutrons using the published value of reaction and scattering cross-sections. The spectra of Po-Be, Pu-Be, Am-Be and Ra-Be arrived at agree well with the published spectra obtained by other methods. Spectrum from spontaneous fission of Cf-252 have also been measured and fitted to the expression N(E)=Esup(1/2)exp(-E/T). The fitted parameter T and spectral details agree well with those in published literature

  5. Combined neutron imaging techniques for cultural heritage purpose

    International Nuclear Information System (INIS)

    Materna, T.

    2009-01-01

    This article presents the different new neutron techniques developed by the Ancient Charm collaboration to image objects of cultural heritage importance: Prompt-gamma-ray activation imaging (PGAI) coupled to cold/thermal neutron transmission tomography, Neutron Resonance Capture Imaging (NRCI) and Neutron Resonance Tomography.

  6. Measurement of the Decay of Thermal Neutrons in Water Poisoned with the Non-1/v Neutron Absorber Cadmium

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, L G; Moeller, E

    1968-01-15

    Measurements have been made of the decay constant of thermal neutrons in water poisoned with the non-1/v absorber cadmium. An experimental method has been used in which proper spatial integration of the neutron flux enables data, representative of the infinite medium to be accumulated without waiting for the establishment of a fundamental mode distribution. The change in effective cross section with concentration of the dissolved cadmium, d{sigma}{sub eff}/dN. has been determined for infinite medium at 20 deg C. Two- and three parameter fits of the decay constant yield -(0.32 {+-} 0.09) x 10{sup -17} barn cm{sup 3} and -(0.47 {+-} 0.10) x 10{sup -17} barn cm{sup 3}, respectively. Earlier published measurements have resulted in two to five times larger values, whereas a published calculated value of Nelkin's model is - 0.33 x 10{sup -17} barn cm{sup 3}.

  7. Measurement of the Decay of Thermal Neutrons in Water Poisoned with the Non-1/v Neutron Absorber Cadmium

    International Nuclear Information System (INIS)

    Larsson, L.G.; Moeller, E.

    1968-01-01

    Measurements have been made of the decay constant of thermal neutrons in water poisoned with the non-1/v absorber cadmium. An experimental method has been used in which proper spatial integration of the neutron flux enables data, representative of the infinite medium to be accumulated without waiting for the establishment of a fundamental mode distribution. The change in effective cross section with concentration of the dissolved cadmium, dσ eff /dN. has been determined for infinite medium at 20 deg C. Two- and three parameter fits of the decay constant yield -(0.32 ± 0.09) x 10 -17 barn cm 3 and -(0.47 ± 0.10) x 10 -17 barn cm 3 , respectively. Earlier published measurements have resulted in two to five times larger values, whereas a published calculated value of Nelkin's model is - 0.33 x 10 -17 barn cm 3

  8. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  9. Fabrication and characterization of dysprosia and alumina based inert matrix neutron absorbers

    International Nuclear Information System (INIS)

    D Ovidio, C.; Oliber, E.; Leiva, S.; Malachevsky, M. T; Taboada, H

    2009-01-01

    Among the elements of the lanthanides series, dysprosium has interesting nuclear properties. Its high thermal neutron absorption cross-section makes it a good neutron absorber. The best ceramic compound apt for nuclear use is its oxide, the disprosia (Dy 2 O 3 ). In order to fabricate neutron absorbers diluted in an inert matrix, it is relevant to study the preparation of a ceramic compound based on alumina (Al 2 O 3 ) and disprosia. In this work, we characterize a particular composition (44,5wt% Dy 2 O 3 , 55,5wt% Al 2 O 3 ) by determining the geometrical density, microstructure and phase formation. The chosen composition corresponds to the lowest temperature eutectic of the alumina-disprosia system, allowing the sintering to proceed at 1700 oC in air. Comparing the data of the green and sinterized pellets, the relative shrinking is of about 17 %, in the same proportion both for diameter and length. The corresponding volumetric reduction is of about 43 %, indicating an increase of the relative geometric density of ∼ 70 %. X-ray diffraction analysis shows the existence of two phases corresponding to the lower eutectic: Dy 3 Al 5 O 1 2 and Al 2 O 3 . The calculated theoretical density is ∼ 5.2 g/cm3. Consequently, the relative density of the pellets is 92 %, indicating the feasibility for the fabrication of the proposed material. In a near future, samples will be irradiated to evaluate their behavior for nuclear use. [es

  10. Cermet based solar selective absorbers : further selectivity improvement and developing new fabrication technique

    OpenAIRE

    Nejati, Mohammadreza

    2008-01-01

    Spectral selectivity of cermet based selective absorbers were increased by inducing surface roughness on the surface of the cermet layer using a roughening technique (deposition on hot substrates) or by micro-structuring the metallic substrates before deposition of the absorber coating using laser and imprint structuring techniques. Cu-Al2O3 cermet absorbers with very rough surfaces and excellent selectivity were obtained by employing a roughness template layer under the infrared reflective l...

  11. More accurate thermal neutron coincidence counting technique

    International Nuclear Information System (INIS)

    Baron, N.

    1978-01-01

    Using passive thermal neutron coincidence counting techniques, the accuracy of nondestructive assays of fertile material can be improved significantly using a two-ring detector. It was shown how the use of a function of the coincidence count rate ring-ratio can provide a detector response rate that is independent of variations in neutron detection efficiency caused by varying sample moderation. Furthermore, the correction for multiplication caused by SF- and (α,n)-neutrons is shown to be separable into the product of a function of the effective mass of 240 Pu (plutonium correction) and a function of the (α,n) reaction probability (matrix correction). The matrix correction is described by a function of the singles count rate ring-ratio. This correction factor is empirically observed to be identical for any combination of PuO 2 powder and matrix materials SiO 2 and MgO because of the similar relation of the (α,n)-Q value and (α,n)-reaction cross section among these matrix nuclei. However the matrix correction expression is expected to be different for matrix materials such as Na, Al, and/or Li. Nevertheless, it should be recognized that for comparison measurements among samples of similar matrix content, it is expected that some function of the singles count rate ring-ratio can be defined to account for variations in the matrix correction due to differences in the intimacy of mixture among the samples. Furthermore the magnitude of this singles count rate ring-ratio serves to identify the contaminant generating the (α,n)-neutrons. Such information is useful in process control

  12. A state-of-the-art report on the development of B{sub 4}C materials as neutron absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Choong Hwan; Kim, Sun Jae; Park, Jee Yun; Kang, Dae Kab [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-01-01

    Boron of 10 atomic weight is one of the best neutron absorbing elements. Among the boron compounds, B{sub 4}C and its composites exhibit excellent material properties. Those materials absorb thermal and fast neutrons, are thermally and chemically very stable, and are very strong in mechanical properties. By neutron irradiation B-10 transforms into Li releasing one He atom. This He release causes swelling, cracking and fragmentation of B{sub 4}C bulks and results in degradation of the materials. The essence of technical developments of B{sub 4}C-based neutron absorbers is the minimization of the effects of He release, and this can be realized through microstructural optimizations of grain and porosity distributions. While pure B{sub 4}C is very difficult in sintering, new neutron absorbing materials of B{sub 4}C-cermets are being developed. B{sub 4}C-cermets are composite materials in which B{sub 4}C powders are dispersed in the metal matrix of Al or Cu. Those materials show easiness in sintering, mechanical forming, and B{sub 4}C content controlling. Neutron absorbing and shielding materials play an important role for the safety of reactor operations and environmental protections. Those materials are being used as monolithic pellets for control rods, burnable poison fuel rods, rack materials for spent fuel storages, shielding materials for shipping casks, and especially for shielding plates for liquid metal reactors. 37 figs., 12 tabs., 41 refs. (Author).

  13. One-speed neutron transport in spheres with totally absorbing cores

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.

    1988-01-01

    Stationary and time-dependent transport of neutrons of one speed has been studied in spheres with totally absorbing cores. For stationary, critical reactors the number of secondaries per collision has been calculated numerically for various inner and outer radii. In the time-dependent case, the decay constant has been calculated for spherical shells of different inner radii and thicknesses. For a fixed ratio between shell thickness and inner radius, the curve of the decay constant versus shell thickness crosses the Corngold limit in the same way as the curve for a homogeneous sphere. When the ratio goes to zero the curve approaches that for an infinite slab. The behaviour is discussed in view of a new result from collision theory, viz. that the following condition must be fulfilled for a body at the point where the decay constant curve crosses the Corngold limit: the average exit distance of the neutrons is equal to the mean free path for scattering

  14. Quantitative determination of absorbed hydrogen in oxidised zircaloy by means of neutron radiography

    International Nuclear Information System (INIS)

    Grosse, M.; Lehmann, E.; Vontobel, P.; Steinbrueck, M.

    2006-01-01

    Hydrogen absorbed in steam-oxidised zircaloy can be determined quantitatively by means of neutron radiography. Correlation parameters between the total cross section and hydrogen content as well as oxide layer thickness were determined quantitatively. At H/Zr atomic ratios lower than 1.0, linear correlations between the hydrogen content and total cross section exist. The total cross section of Zr is lower and the effect of the hydrogen is higher in radiography measurements with a cold neutron spectrum than with a thermal spectrum. A Be filter reduces the effects of lower wavelength and epithermal neutrons and extends the linear correlations to higher H/Zr atomic ratios. Due to the better possibilities of background corrections, the neutron image should be detected by a CCD camera for a proper quantitative analysis with a medium spatial resolution of about 0.1 mm. A higher spatial resolution, but larger uncertainties in the quantitative hydrogen determination are achieved by measurements with imaging plates. The effect of oxygen layers on the total cross section is much smaller than the effect of hydrogen. The total cross section measured depends linearly on the oxide layer thickness

  15. Effect of absorption discontinuity on neutron spectra of water assemblies poisoned with non-1/V absorbers

    International Nuclear Information System (INIS)

    Gupta, I.J.; Trikha, S.K.

    1977-01-01

    Calculations are presented of the diffusion of thermal neutrons (2.5 x 10 -4 to 7 x 10 -1 eV) across an absorption discontinuity in a water assembly, consisting of pure water on one side and aqueous solutions of three different non-1/V absorbers on the other, which were obtained by solving the Boltzmann transport equation in the diffusion approximation using the multigroup formalism. The gradual appearance and disappearance of the depletion region in the neutron spectra (caused by the resonance absorption peaks at energies 0.096 and 0.179 eV for samarium and cadmium respectively), as one moves from the pure water assembly to the poisoned water assembly and vice versa, have also been studied. The minimum concentrations of Sm and Cd atoms in water for which the depletion region in the spectra just starts building up are found to be 60 x 10 18 Sm atom cm -3 and 125 x 10 18 Cd atom cm -3 respectively. However no such depletion region is observed in gadolinium-poisoned water assembly. At the boundary, the equilibrium neutron distribution gets disturbed and is re-established to the equilibrium distribution of the second medium at some distance from the interface. The diffusion lengths so calculated from the total neutron density curves are in good agreement with the experimental results of Goddard and Johnson (Nucl. Sci. Eng.; 37:127 (1969)) at various concentrations of Gd and Cd atoms in water. (author)

  16. Absorbed bone marrow dose in certain dental radiographic techniques

    International Nuclear Information System (INIS)

    White, S.C.; Rose, T.C.

    1979-01-01

    The absorbed dose of radiation in the bone marrow of the region of the head and neck was measured during intraoral, panoramic, and cephalometric radiography. Panoramic radiography results in a dose a fifth or less than that from an intraoral survey. The use of rectangular collimation reduces the bone marrow absorbed dose from an intraoral survey by about 60%. Comparison of the doses from dental radiography with natural environmental radiation shows that an intraoral set of films results in the same total dose to the bone marrow as 65 days of background exposure. The use of rectangular collimation reduces this value to 25 days. Panoramic radiography results in significantly less irradiation, as it reduces the value to 14 days or fewer. Dental radiography thus involves exposures in the range of variation of natural environmental background values

  17. Lectures on neutron scattering techniques: 1

    International Nuclear Information System (INIS)

    Carlile, C.J.

    1988-08-01

    The lecture on the production of neutrons was presented at a Summer School on neutron scattering, Rome, 1986. A description is given of the production of neutrons by natural radioactive sources, fission, and particle accelerator sources. Modern neutron sources with high intensities are discussed including the ISIS pulsed neutron source at the Rutherford Appleton Laboratory and the High Flux Reactor at the Institut Laue Langevin. (U.K.)

  18. Preparation and characterization of ceramic neutron absorbers based on dysprosia and gadolinia

    International Nuclear Information System (INIS)

    Burgos, F.; Oliber, E.; Leiva S; Lestani, H.; Malachevsky, M.T.; Taboada, H.; D'Ovidio, C.

    2012-01-01

    Among the elements of the lanthanide series, dysprosium and gadolinium have interesting nuclear properties. Due to their high thermal neutron absorption cross-section they are good neutron absorbers. The only compounds suitable for nuclear use are their oxides, dysprosia (Dy 2 O 3 ) and gadolinia (Gd 2 O 3 ). To fabricate neutron absorbers diluted in an inert matrix, e.g. alumina (Al 2 O 3 ), it is relevant to study the preparation of a ceramic compound based on alumina (Al 2 O 3 ) and dysprosia or gadolinia. In this work, we characterize four different nominal compositions with high contents of gadolinia and dysprosia: (a) (45 wt% Dy 2 O 3 , 55 wt% Al 2 O 3 ), (b) (93 wt% Dy 2 O 3 , 7 wt% Al 2 O 3 ), (c) (50 wt% Gd 2 O 3 , 50 wt% Al 2 O 3 ) and (d) (90 wt% Gd 2 O 3 , 10 wt% Al 2 O 3 ). These compositions were selected as their stoichiometry correspond to the eutectic phases found in the respective phase diagrams, so as to attain sinterization at lower temperatures of approximately 1700 o C in air. The investigated parameters are the geometrical density of the pellets, the microstructure and the phases observed using x-ray diffraction. Contraction of the pellets was obtained by measuring the volumetric change between the green and the sintered samples. It was observed that the relative contraction was the same both in thickness and diameter. We discuss the eutectic phase formation and densification observed for the different compositions (author)

  19. Neutron logging reliability techniques and apparatus

    International Nuclear Information System (INIS)

    Johnstone, C.W.

    1975-01-01

    Apparatus and methods for verifying the validity of data derived at least in part by neutron logging of earth formations, and, where indicated, for affording neutron diffusion-corrected values of such data, are disclosed. (WHK)

  20. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

    1987-12-01

    This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

  1. Response functions for computing absorbed dose to skeletal tissues from neutron irradiation

    Science.gov (United States)

    Bahadori, Amir A.; Johnson, Perry; Jokisch, Derek W.; Eckerman, Keith F.; Bolch, Wesley E.

    2011-11-01

    Spongiosa in the adult human skeleton consists of three tissues—active marrow (AM), inactive marrow (IM) and trabecularized mineral bone (TB). AM is considered to be the target tissue for assessment of both long-term leukemia risk and acute marrow toxicity following radiation exposure. The total shallow marrow (TM50), defined as all tissues lying within the first 50 µm of the bone surfaces, is considered to be the radiation target tissue of relevance for radiogenic bone cancer induction. For irradiation by sources external to the body, kerma to homogeneous spongiosa has been used as a surrogate for absorbed dose to both of these tissues, as direct dose calculations are not possible using computational phantoms with homogenized spongiosa. Recent micro-CT imaging of a 40 year old male cadaver has allowed for the accurate modeling of the fine microscopic structure of spongiosa in many regions of the adult skeleton (Hough et al 2011 Phys. Med. Biol. 56 2309-46). This microstructure, along with associated masses and tissue compositions, was used to compute specific absorbed fraction (SAF) values for protons originating in axial and appendicular bone sites (Jokisch et al 2011 Phys. Med. Biol. 56 6857-72). These proton SAFs, bone masses, tissue compositions and proton production cross sections, were subsequently used to construct neutron dose-response functions (DRFs) for both AM and TM50 targets in each bone of the reference adult male. Kerma conditions were assumed for other resultant charged particles. For comparison, AM, TM50 and spongiosa kerma coefficients were also calculated. At low incident neutron energies, AM kerma coefficients for neutrons correlate well with values of the AM DRF, while total marrow (TM) kerma coefficients correlate well with values of the TM50 DRF. At high incident neutron energies, all kerma coefficients and DRFs tend to converge as charged-particle equilibrium is established across the bone site. In the range of 10 eV to 100 Me

  2. Discussions in symposium 'neutron dosimetry in neutron fields - from detection techniques to medical applications'

    International Nuclear Information System (INIS)

    Tanimura, Y.; Sato, T.; Kumada, H.; Terunuma, T.; Sakae, T.; Harano, H.; Matsumoto, T.; Suzuki, T.; Matsufuji, N.

    2008-01-01

    Recently the traceability system (JCSS) of neutron standard based on the Japanese law 'Measurement Act' has been instituted. In addition, importance of the neutron dose evaluation has been increasing in not only the neutron capture medical treatment but also the proton or heavy particle therapy. Against such a background, a symposium 'Neutron dosimetry in neutron fields - From detection techniques to medical applications-' was held on March 29, 2008 and recent topics on the measuring instruments and their calibration, the traceability system, the simulation technique and the medical applications were introduced. This article summarizes the key points in the discussion at the symposium. (author)

  3. Calibration of the JET neutron yield monitors using the delayed neutron counting technique

    International Nuclear Information System (INIS)

    van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.

    1990-01-01

    The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET

  4. Enriched boric acid as an optimized neutron absorber in the EPR primary coolant

    International Nuclear Information System (INIS)

    Cosse, Christelle; Jolivel, Fabienne; Berger, Martial

    2012-09-01

    This paper focuses on one of the most important EPR PWR reactor design optimizations, through primary coolant conditioning by enriched boric acid (EBA). On PWRs throughout the world, boric acid has already been implemented in primary coolant and associated auxiliary systems for criticality control, due to its high Boron 10 neutron absorption cross section. Boric acid also allows primary coolant pH 300C control in combination with lithium hydroxide in many PWRs. The boric acid employed in the majority of existing PWRs is the 'natural' one, with a typical isotopic atomic abundance in Boron 10 about 19.8 at.%. However, EPR requirements for neutron management are more important, due to its fully optimized design compared to older PWRs. From the boron point of view, it means that criticality could be controlled either by increased 'natural' Boron concentrations or by using EBA. Comparatively to 'natural' boric acid, EBA allows for: - the use of smaller storage volumes for an identical total Boron concentration, or lower total Boron concentration if the tank volumes are kept identical. The latter also reduces the risks of boric acid crystallization, in spite of increased neutron-absorbing properties - the application of an evolutionary chemistry operating regime called Advanced pH Control, making it possible to maintain a constant pH 300C value at 7.2 in the primary coolant at nominal conditions throughout entire cycles. This optimized stability of pH 300C will contribute to reduce the consequences of contamination of the reactor coolant system by corrosion products, and consequently, all related issues - the reduction of borated liquid wastes, thanks to maximal recycling resulting from EPR design. The increased design costs associated with EBA are consequently compensated by a reduced total consumption of this chemical. Therefore, the basic design choice for the EPR is the use of EBA. For the Flamanville 3 EPR, according to the above

  5. Application of the pulsed neutron technique on the reactors ALIZE - AQUILON (1963); Application de la methode des neutrons pulses sur les piles ALIZE et AQUILON (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Jacquemart, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    Different methods of measuring the ratio effective delayed fraction / prompt neutron lifetime, {alpha}{sub c}, are described. According to the classic pulsed neutron technique the negative reactivity due to a localized absorber is given by {rho} / {beta}{sub eff} = {alpha} / {alpha}{sub c} -1 Experiments are reported which show that in this case {alpha}{sub c} can not be considered constant for large reactivities. The absorber element distorts the flux in the system, increasing the importance of the reflector. An application of the pulsed neutron method to the measurement of critical distributed boron concentrations of various absorber elements is described. Less time is required than for the usual super-critical techniques, and the experimental analysis is simplified. It is interesting to note that the results are not influenced by the spectral sensitivity of the control element. A modified pulsed neutron method has been tried out. This procedure was used to determine by measurements at sub-critical the critical water level of uranium-heavy water lattices with a high precision. (author) [French] Differents modes operatoires pour definir la valeur du rapport pourcentage effectif de neutrons retardes / temps de vie, {alpha}{sub c}, sont exposes. La methode classique par neutrons pulses definit l'anti-reactivite d'un element absorbant a partir de la relation: {rho} / {beta}{sub eff} {alpha} / {alpha}{sub c} -1 Les manipulations effectuees montrent qu'on ne peut considerer dans ce cas {alpha}{sub c} constant pour de tres grandes anti-reactivites. L'absorbant introduit dans la pile deforme le flux et augmente l'importance du reflecteur. Une application de la methode des neutrons pulses pour mesurer le titre critique en mg de B/l de divers absorbants est signalee. Les operations sont effectuees en regime sous-critique avec un certain gain de temps et une grande facilite de depouillement. Il est interessant de noter que les resultats ne sont pas

  6. Application of the pulsed neutron technique on the reactors ALIZE - AQUILON (1963); Application de la methode des neutrons pulses sur les piles ALIZE et AQUILON (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Jacquemart, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    Different methods of measuring the ratio effective delayed fraction / prompt neutron lifetime, {alpha}{sub c}, are described. According to the classic pulsed neutron technique the negative reactivity due to a localized absorber is given by {rho} / {beta}{sub eff} = {alpha} / {alpha}{sub c} -1 Experiments are reported which show that in this case {alpha}{sub c} can not be considered constant for large reactivities. The absorber element distorts the flux in the system, increasing the importance of the reflector. An application of the pulsed neutron method to the measurement of critical distributed boron concentrations of various absorber elements is described. Less time is required than for the usual super-critical techniques, and the experimental analysis is simplified. It is interesting to note that the results are not influenced by the spectral sensitivity of the control element. A modified pulsed neutron method has been tried out. This procedure was used to determine by measurements at sub-critical the critical water level of uranium-heavy water lattices with a high precision. (author) [French] Differents modes operatoires pour definir la valeur du rapport pourcentage effectif de neutrons retardes / temps de vie, {alpha}{sub c}, sont exposes. La methode classique par neutrons pulses definit l'anti-reactivite d'un element absorbant a partir de la relation: {rho} / {beta}{sub eff} {alpha} / {alpha}{sub c} -1 Les manipulations effectuees montrent qu'on ne peut considerer dans ce cas {alpha}{sub c} constant pour de tres grandes anti-reactivites. L'absorbant introduit dans la pile deforme le flux et augmente l'importance du reflecteur. Une application de la methode des neutrons pulses pour mesurer le titre critique en mg de B/l de divers absorbants est signalee. Les operations sont effectuees en regime sous-critique avec un certain gain de temps et une grande facilite de depouillement. Il est interessant de noter que les resultats ne sont pas affectes par la

  7. Absorbant materials

    International Nuclear Information System (INIS)

    Quetier, Monique.

    1978-11-01

    Absorbants play a very important part in the nuclear industry. They serve for the control, shut-down and neutron shielding of reactors and increase the capacity of spent fuel storage pools and of special transport containers. This paper surveys the usual absorbant materials, means of obtainment, their essential characteristics relating to their use and their behaviour under neutron irradiation [fr

  8. Neutron logging reliability techniques and apparatus

    International Nuclear Information System (INIS)

    Johnstone, C.W.

    1978-01-01

    This invention relates in general to neutron logging of earth formations, and in particular, to novel apparatus and procedures for determining the validity, or reliability, of data derived at least in part by logging neutron characteristics of earth formations and, if desired, for affording verifiably accurate indications of such data

  9. Neutron logging reliability techniques and apparatus

    International Nuclear Information System (INIS)

    Johnstone, C.W.

    1974-01-01

    This invention relates in general to neutron logging of earth formations, and in particular, to novel apparatus and procedures for determining the validity, or reliability, of data derived at least in part by logging neutron characteristics of earth formations and, if desired, for affording verifiably accurate indications of such data. (author)

  10. Techniques of in vivo neutron activation analysis

    International Nuclear Information System (INIS)

    Chettle, D.R.; Fremlin, J.H.

    1984-01-01

    This review is dealt with under the following headings, intended to reflect the different factors affecting the measurement sensitivity, starting with the choice of neutron source and proceeding, through the reaction characteristics, to the detection system, the questions of dosimetry and ethical constraints being also discussed: 1) neutron sources, slowing down and interaction processes, energy spectrum and flux uniformity, timing 2) neutron reactions used for in vivo analyses 3) detectors, choice, geometrical considerations and detector shielding 4) data collection and processing 5) interpretation, major elements, absolute or sequential measurements, relationship to other parameters 6) dosimetry, framework for dose levels, biological effects of neutron interactions, neutron doses in practice 7) implications for measurement of calcium, nitrogen and cadmium. (U.K.)

  11. Investigation of neutron guide systems: Analysis techniques and an experiment

    International Nuclear Information System (INIS)

    Kudryashev, V.A.

    1991-01-01

    This paper discusses the in-depth study of the specific characteristics of the physical processes associated with the total reflection of neutrons from actual reflective coatings; the study of the process whereby neutrons transit a nonideal image channel with allowance for the aforementioned characteristics, and; the development of physical criteria and techniques for calculating the optimum geometry of a neutron guide source system based on the laws found to govern this transit process

  12. Evaluation of neutron techniques for illicit substance detection

    International Nuclear Information System (INIS)

    Fink, C.L.; Micklich, B.J.; Yule, T.J.; Humm, P.; Sagalovsky, L.; Martin, M.M.

    1995-01-01

    We are studying inspection systems based on the use of fast neutrons for detecting illicit substances such as explosives and drugs in luggage and cargo containers. Fast-neutron techniques can determine the quantities of light elements such as carbon, nitrogen, and oxygen in a volume element. Illicit substances containing these elements are characterized by distinctive elemental densities or density ratios. We discuss modeling and tomographic reconstruction studies for fast-neutron transmission spectroscopy. (orig.)

  13. Evaluation of neutron techniques for illicit substance detection

    International Nuclear Information System (INIS)

    Fink, C.L.; Micklich, B.J.; Yule, T.J.; Humm, P.; Sagalovsky, L.; Martin, M.M.

    1994-01-01

    The authors are studying inspection systems based on the use of fast neutrons for detecting illicit substances such as explosives and drugs in luggage and cargo containers. Fast neutron techniques can determine the quantities of light elements such as carbon, nitrogen, and oxygen in a volume element. Illicit substances containing these elements are characterized by distinctive elemental densities or density ratios. They discuss modeling and tomographic reconstruction studies for fast-neutron transmission spectroscopy

  14. Experimental techniques for the consolidation of the neutron spectrum

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Unfolding techniques are widely known but their use is not widespread due to their complexity. Such procedure consists in the adjustment of calculated quantities to experimental results by the modification of the neutron spectrum, getting correction factors for the calculated quantities. In this work we describe the general procedure that must be executed for a neutron spectrum unfolding. (author) [es

  15. Neutron therapy coupling brachytherapy and boron neutron capture therapy (BNCT) techniques

    International Nuclear Information System (INIS)

    Chaves, Iara Ferreira.

    1994-12-01

    In the present dissertation, neutron radiation techniques applied into organs of the human body are investigated as oncologic radiation therapy. The proposal treatment consists on connecting two distinct techniques: Boron Neutron Capture Therapy (BNCT) and irradiation by discrete sources of neutrons, through the brachytherapy conception. Biological and radio-dosimetrical aspects of the two techniques are considered. Nuclear aspects are discussed, presenting the nuclear reactions occurred in tumoral region, and describing the forms of evaluating the dose curves. Methods for estimating radiation transmission are reviewed through the solution of the neutron transport equation, Monte Carlo methodology, and simplified analytical calculation based on diffusion equation and numerical integration. The last is computational developed and presented as a quickly way to neutron transport evaluation in homogeneous medium. The computational evaluation of the doses for distinct hypothetical situations is presented, applying the coupled techniques BNTC and brachytherapy as an possible oncologic treatment. (author). 78 refs., 61 figs., 21 tabs

  16. Pulsed neutron generator for mass flow measurement using the pulsed neutron activation technique

    International Nuclear Information System (INIS)

    Rochau, G.E.; Hornsby, D.R.; Mareda, J.F.; Riggan, W.C.

    1980-01-01

    A high-output, transportable neutron generator has been developed to measure mass flow velocities in reactor safety tests using the Pulsed Neutron Activation (PNA) Technique. The PNA generator produces >10 10 14 MeV D-T neutrons in a 1.2 millisecond pulse. The Millisecond Pulse (MSP) Neutron Tube, developed for this application, has an expected operational life of 1000 pulses, and it limits the generator pulse repetition rate to 12 pulses/minute. A semiconductor neutron detector is included in the generator package to monitor the neutron output. The control unit, which can be operated manually or remotely, also contains a digital display with a BCD output for the neutron monitor information. The digital logic of the unit controls the safety interlocks and rejects transient signals which could accidently fire the generator

  17. Evaluation of neutron dosimetry techniques for well-logging operations

    International Nuclear Information System (INIS)

    Cummings, F.M.; Haggard, D.L.; Endres, G.W.R.

    1985-07-01

    Neutron dose and energy spectral measurements from 241 AmBe and a 14 MeV neutron generator were performed at a well-logging laboratory. The measurement technique included the tissue equivalent proportional counter, multisphere, two types of remmeters and five types of personnel neutron dosimeters. Several source configurations were used to attempt to relate data to field situations. The results of the measurements indicated that the thermoluminescent albedo dosimeter was the most appropriate personnel neutron dosimeter, and that the most appropriate calibration source would be the source normally employed in the field with the calibration source being used in the unmoderated configuration. 7 refs., 35 figs., 14 tabs

  18. Application of neutron backscatter techniques to level measurement problems

    International Nuclear Information System (INIS)

    Leonardi-Cattolica, A.M.; McMillan, D.H.; Telfer, A.; Griffin, L.H.; Hunt, R.H.

    1982-01-01

    We have designed and built portable level detectors and fixed level monitors based on neutron scattering and detection principles. The main components of these devices, which we call neutron backscatter gauges, are a neutron emitting radioisotope, a neutron detector, and a ratemeter. The gauge is a good detector for hydrogen but is much less sensitive to most other materials. This allows level measurements of hydrogen bearing materials, such as hydrocarbons, to be made through the walls of metal vessels. Measurements can be made conveniently through steel walls which are a few inches thick. We have used neutron backscatter gauges in a wide variety of level measurement applications encountered in the petrochemical industry. In a number of cases, the neutron techniques have proven to be superior to conventional level measurement methods, including gamma ray methods

  19. Elements of slow-neutron scattering basics, techniques, and applications

    CERN Document Server

    Carpenter, J M

    2015-01-01

    Providing a comprehensive and up-to-date introduction to the theory and applications of slow-neutron scattering, this detailed book equips readers with the fundamental principles of neutron studies, including the background and evolving development of neutron sources, facility design, neutron scattering instrumentation and techniques, and applications in materials phenomena. Drawing on the authors' extensive experience in this field, this text explores the implications of slow-neutron research in greater depth and breadth than ever before in an accessible yet rigorous manner suitable for both students and researchers in the fields of physics, biology, and materials engineering. Through pedagogical examples and in-depth discussion, readers will be able to grasp the full scope of the field of neutron scattering, from theoretical background through to practical, scientific applications.

  20. Calibration technique for the neutron surface moisture measurement system

    International Nuclear Information System (INIS)

    Watson, W.T.; Shreve, D.C.

    1996-01-01

    A technique for calibrating the response of a surface neutron moisture measurement probe to material moisture concentration has been devised. Tests to ensure that the probe will function in the expected in-tank operating environment are also outlined

  1. Neutron activation analysis: an emerging technique for conservation/preservation

    International Nuclear Information System (INIS)

    Sayre, E.V.

    1976-01-01

    The diverse applications of neutron activation in analysis, preservation, and documentation of art works and artifacts are described with illustrations for each application. The uses of this technique to solve problems of attribution and authentication, to reveal the inner structure and composition of art objects, and, in some instances to recreate details of the objects are described. A brief discussion of the theory and techniques of neutron activation analysis is also included

  2. Experimental technique of small angle neutron scattering

    International Nuclear Information System (INIS)

    Xia Qingzhong; Chen Bo

    2006-03-01

    The main parts of Small Angle Neutron Scattering (SANS) spectrometer, and their function and different parameters are introduced from experimental aspect. Detailed information is also introduced for SANS spectrometer 'Membrana-2'. Based on practical experiments, the fundamental requirements and working condition for SANS experiments, including sample preparation, detector calibration, standard sample selection and data preliminary process are described. (authors)

  3. Neutron spin echo: A new concept in polarized thermal neutron techniques

    International Nuclear Information System (INIS)

    Mezei, F.

    1980-01-01

    A simple method to change and keep track of neutron beam polarization non-parallel to the magnetic field is described. It makes possible the establishment of a new focusing effect we call neutron spin echo. The technique developed and tested experimentally can be applied in several novel ways, e.g. for neutron spin flipper of superior characteristics, for a very high resolution spectrometer for direct determination of the Fourier transform of the scattering function, for generalised polarization analysis and for the measurement of neutron particle properties with significantly improved precision. (orig.)

  4. Slow neutron mapping technique for level interface measurement

    Science.gov (United States)

    Zain, R. M.; Ithnin, H.; Razali, A. M.; Yusof, N. H. M.; Mustapha, I.; Yahya, R.; Othman, N.; Rahman, M. F. A.

    2017-01-01

    Modern industrial plant operations often require accurate level measurement of process liquids in production and storage vessels. A variety of advanced level indicators are commercially available to meet the demand, but these may not suit specific need of situations. The neutron backscatter technique is exceptionally useful for occasional and routine determination, particularly in situations such as pressure vessel with wall thickness up to 10 cm, toxic and corrosive chemical in sealed containers, liquid petroleum gas storage vessels. In level measurement, high energy neutrons from 241Am-Be radioactive source are beamed onto a vessel. Fast neutrons are slowed down mostly by collision with hydrogen atoms of material inside the vessel. Parts of thermal neutron are bounced back towards the source. By placing a thermal detector next to the source, these backscatter neutrons can be measured. The number of backscattered neutrons is directly proportional to the concentration of the hydrogen atoms in front of the neutron detector. As the source and detector moved by the matrix around the side of the vessel, interfaces can be determined as long as it involves a change in hydrogen atom concentration. This paper presents the slow neutron mapping technique to indicate level interface of a test vessel.

  5. Steel research using neutron beam techniques. In-situ neutron diffraction, small-angle neutron scattering and residual stress analysis

    International Nuclear Information System (INIS)

    Sueyoshi, Hitoshi; Ishikawa, Nobuyuki; Yamada, Katsumi; Sato, Kaoru; Nakagaito, Tatsuya; Matsuda, Hiroshi; Arakaki, Yu; Tomota, Yo

    2014-01-01

    Recently, the neutron beam techniques have been applied for steel researches and industrial applications. In particular, the neutron diffraction is a powerful non-destructive method that can analyze phase transformation and residual stress inside the steel. The small-angle neutron scattering is also an effective method for the quantitative evaluation of microstructures inside the steel. In this study, in-situ neutron diffraction measurements during tensile test and heat treatment were conducted in order to investigate the deformation and transformation behaviors of TRIP steels. The small-angle neutron scattering measurements of TRIP steels were also conducted. Then, the neutron diffraction analysis was conducted on the high strength steel weld joint in order to investigate the effect of the residual stress distribution on the weld cracking. (author)

  6. Measurement of epithermal neutrons by a coherent demodulation technique

    CERN Document Server

    Horiuchi, N; Takahashi, H; Kobayashi, H; Harasawa, S

    2000-01-01

    Epithermal neutrons have been measured using a neutron dosimeter via a coherent demodulation technique. This dosimeter consists of CsI(Tl)-photodiode scintillation detectors, four of which are coupled to neutron-gamma converting foils of various sizes. Neutron-gamma converting foils of In, Au and Co materials were used, each of which has a large capture cross section which peaks in the epithermal neutron energy region. The type of foil was selected according to the material properties that best correspond to the energy of the epithermal neutrons to be measured. In addition, the proposed technique was applied using Au-foils in order to measure the Cd ratio. The validity of the proposed technique was examined using an sup 2 sup 4 sup 1 Am-Be source placed in a testing stack of polyethylene blocks, and the results were compared with the theoretical values calculated by the Monte Carlo calculation. Finally, the dosimeter was applied for measuring epithermal neutrons and the Cd ratio in an experimental beam-tube o...

  7. Identification of Fissionable Materials Using the Tagged Neutron Technique

    International Nuclear Information System (INIS)

    Keegan, R.P.; Hurley, J.P.; Tinsley, J.R.; Trainham, R.

    2009-01-01

    This summary describes experiments to detect and identify fissionable materials using the tagged neutron technique. The objective of this work is to enhance homeland security capability to find fissionable material that may be smuggled inside shipping boxes, containers, or vehicles. The technique distinguishes depleted uranium from lead, steel, and tungsten. Future work involves optimizing the technique to increase the count rate by many orders of magnitude and to build in the additional capability to image hidden fissionable materials. The tagged neutron approach is very different to other techniques based on neutron die-away or photo-fission. This work builds on the development of the Associated Particle Imaging (API) technique at the Special Technologies Laboratory (STL). Similar investigations have been performed by teams at the Oak Ridge National Laboratory (ORNL), the Khlopin Radium Institute in Russia, and by the EURITRACK collaboration in the European Union

  8. Standard specification for boron-Based neutron absorbing material systems for use in nuclear spent fuel storage racks

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 This specification defines criteria for boron-based neutron absorbing material systems used in racks in a pool environment for storage of nuclear light water reactor (LWR) spent-fuel assemblies or disassembled components to maintain sub-criticality in the storage rack system. 1.2 Boron-based neutron absorbing material systems normally consist of metallic boron or a chemical compound containing boron (for example, boron carbide, B4C) supported by a matrix of aluminum, steel, or other materials. 1.3 In a boron-based absorber, neutron absorption occurs primarily by the boron-10 isotope that is present in natural boron to the extent of 18.3 ± 0.2 % by weight (depending upon the geological origin of the boron). Boron, enriched in boron-10 could also be used. 1.4 The materials systems described herein shall be functional – that is always be capable to maintain a B10 areal density such that subcriticality Keff <0.95 or Keff <0.98 or Keff < 1.0 depending on the design specification for the service...

  9. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  10. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Wang, Lumin; Wierschke, Jonathan Brett

    2015-01-01

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H 3 BO 3 ). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  11. Determination of contraband using fast neutron resonance technique

    Energy Technology Data Exchange (ETDEWEB)

    Bae, J.; Whang, J. [Kyunghee Univ., Dept. of Nuclear Engineering, Yongin-shi, Kyongki-do (Korea, Republic of)

    2004-07-01

    'Full-text:' Resonance technique with monoenergetic fast neutron beam is able to map features in bulk samples in a way that is sensitive to their elemental composition. It has a number of potential applications, for example, in mining and in the detection of contraband materials such as illicit drugs and explosives. By moving around the neutron detector experiences neutrons in the form of narrow line beam with different energies as the angle to the neutron source changes. Projection data was obtained using the Monte Carlo code MCNP4C. Therefore the fast neutrons scattered from an unknown object are used to determine the elemental content of the object and hence lead to its identification. Scattered features simulated for various test materials are analyzed using the HEPRO program system (PTB, Braunschweig) to determine the atom weight fractions for H. C. N, O and other elements in the materials. Atom weight fractions determined from scattering features are insensitive to neutron interactions in interfering materials surrounding the object. The simulations demonstrate that the fast neutron resonance technique (FNRT) provides reliable elemental characterization of bulk materials and has the necessary sensitivity to distinguish between drugs, explosives and other materials. (author)

  12. Determination of contraband using fast neutron resonance technique

    International Nuclear Information System (INIS)

    Bae, J.; Whang, J.

    2004-01-01

    'Full-text:' Resonance technique with monoenergetic fast neutron beam is able to map features in bulk samples in a way that is sensitive to their elemental composition. It has a number of potential applications, for example, in mining and in the detection of contraband materials such as illicit drugs and explosives. By moving around the neutron detector experiences neutrons in the form of narrow line beam with different energies as the angle to the neutron source changes. Projection data was obtained using the Monte Carlo code MCNP4C. Therefore the fast neutrons scattered from an unknown object are used to determine the elemental content of the object and hence lead to its identification. Scattered features simulated for various test materials are analyzed using the HEPRO program system (PTB, Braunschweig) to determine the atom weight fractions for H. C. N, O and other elements in the materials. Atom weight fractions determined from scattering features are insensitive to neutron interactions in interfering materials surrounding the object. The simulations demonstrate that the fast neutron resonance technique (FNRT) provides reliable elemental characterization of bulk materials and has the necessary sensitivity to distinguish between drugs, explosives and other materials. (author)

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

    International Nuclear Information System (INIS)

    Sekine, T.; Baba, H.

    1978-01-01

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

  14. A neutron calibration technique for detectors with low neutron/high photon sensitivity

    International Nuclear Information System (INIS)

    Jahr, R.; Guldbakke, S.; Cosack, M.; Dietze, G.; Klein, H.

    1978-03-01

    The neutron response of a detector with low neutron-/high photon sensitivity is given by the difference of two terms: the response to the mixed neutron-photon field, measured directly, and the response to the photons, deduced from additional measurements with a photon spectrometer. The technique is particularly suited for use in connection with targets which consist of a thick backing and thin layer of neutron producing material such as T, D, Li nuclei. Then the photon component of the mixed field is very nearly the same as the pure photon field from a 'phantom target', being identical with the neutron producing target except for the missing neutron producing material. Using this technique in connection with a T target (Ti-T-layer on silver backing) and the corresponding phantom target (Ti-layer on silver backing), a GM counter was calibrated at a neutron energy of 2.5 MeV. Possibilities are discussed to subsequently calibrate the GM counter at other neutron energies without the use of the photon spectrometer. (orig./HP) [de

  15. A new bridge technique for neutron tomography and diffraction measurements

    International Nuclear Information System (INIS)

    Burca, G.; James, J.A.; Kockelmann, W.; Fitzpatrick, M.E.; Zhang, S.Y.; Hovind, J.; Langh, R. van

    2011-01-01

    An attractive feature of neutron techniques is the ability to identify hidden materials and structures inside engineering components and objects of art and archaeology. Bearing this in mind we are investigating a new technique, 'Tomography Driven Diffraction' (TDD), that exploits tomography data to guide diffraction experiments on samples with complex structures and shapes. The technique can be used utilising combinations of individual tomography and diffraction instruments, such as NEUTRA (PSI, CH) and ENGIN-X (ISIS, UK), but is also suitable for new combined imaging and diffraction instruments such as the JEEP synchrotron engineering instrument (DIAMOND, UK) and the proposed IMAT neutron imaging and diffraction instrument (ISIS, UK).

  16. First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

    Science.gov (United States)

    You, Yan; Yoshida, Katsumi; Yano, Toyohiko

    2018-05-01

    Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

  17. X-ray and neutron techniques for nanomaterials characterization

    CERN Document Server

    2016-01-01

    Fifth volume of a 40 volume series on nanoscience and nanotechnology, edited by the renowned scientist Challa S.S.R. Kumar. This handbook gives a comprehensive overview about X-ray and Neutron Techniques for Nanomaterials Characterization. Modern applications and state-of-the-art techniques are covered and make this volume an essential reading for research scientists in academia and industry.

  18. Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  19. Elemental analysis of brazing alloy samples by neutron activation technique

    International Nuclear Information System (INIS)

    Eissa, E.A.; Rofail, N.B.; Hassan, A.M.; El-Shershaby, A.; Walley El-Dine, N.

    1996-01-01

    Two brazing alloy samples (C P 2 and C P 3 ) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10 1 1 n/cm 2 /s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10 1 2 n/cm 2 /s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab

  20. Elemental analysis of brazing alloy samples by neutron activation technique

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, E A; Rofail, N B; Hassan, A M [Reactor and Neutron physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); El-Shershaby, A; Walley El-Dine, N [Physics Department, Faculty of Girls, Ain Shams Universty, Cairo (Egypt)

    1997-12-31

    Two brazing alloy samples (C P{sup 2} and C P{sup 3}) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10{sup 1}1 n/cm{sup 2}/s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10{sup 1}2 n/cm{sup 2}/s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab.

  1. Nuclear criticality safety: general. 6. Application of Fixed Neutron Absorbers in the New Hanford PFP Horizontal Rack Design

    International Nuclear Information System (INIS)

    Lan, J.S.; Miller, E.M.; Toffer, H.; Mo, B.S.

    2001-01-01

    The Hanford Plutonium Finishing Plant (PFP) is currently in a waste cleanup and plutonium stabilization mode. Plutonium-bearing materials are processed through thermal treatment, creating forms of oxides suitable for long-term storage. Stabilized materials at PFP are stored in a variety of cans such as the bag-less transfer cans (BTCs), which are ultimately contained in the U.S. Department of Energy (DOE) 3013 can; both cans are larger than previously used plutonium storage containers and hold more plutonium. To compensate for the increased plutonium loadings, added engineered safety features were considered in the storage facilities. The vaults in PFP, subdivided into concrete-walled cubicles, will contain both new and older cans. The DOE 3013 and BTC cans may be loaded with up to 4.4 kg of plutonium as a compound (mostly oxide). New racks that store cans horizontally are being constructed to hold both new and older containers. The loading objective is to accommodate 70 kg of plutonium per cubicle. Two design analysis approaches for the new racks were considered. The first approach incorporated neutron absorption provided by the structural materials of the rack and the cans in determining a safe configuration. A rack loading arrangement was determined as shown in Fig. 1 and specified in Table I. This approach provides compliance with criticality control requirements; however, added administrative controls were needed to accommodate a sufficient number of cans in specific locations to achieve 70 kg of plutonium per cubicle. The 4.4-kg plutonium container can be placed only in predetermined locations. The second approach evaluated the addition of a fixed neutron absorber plate along the back wall of the cubicle (Fig. 1). The location of the special plate facilitates installation of the racks and provides additional criticality safety margin beyond the first approach. Its presence permits loading of racks with up to 4.4-kg plutonium cans in any storage locations

  2. Burnable absorber-integrated Guide Thimble (BigT) - 1. Design concepts and neutronic characterization on the fuel assembly benchmarks

    International Nuclear Information System (INIS)

    Yahya, Mohd-Syukri; Yu, Hwanyeal; Kim, Yonghee

    2016-01-01

    This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named 'Burnable absorber-integrated Guide Thimble' (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal-hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17x17 and 16x16 fuel assembly lattices. For the 17x17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16x16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library. (author)

  3. . Estimating soil contamination from oil spill using neutron backscattering technique

    International Nuclear Information System (INIS)

    Okunade, I.O.; Jonah, S.A.; Abdulsalam, M.O.

    2009-01-01

    An analytical facility which is based on neutron backscattering technique has been adapted for monitoring oil spill. The facility which consists of 1 Ci Am-Be isotopic source and 3 He neutron detector is based on the principle of slowing down of neutrons in a given medium which is dominated by the elastic process with the hydrogen nucleus. Based on this principle, the neutron reflection parameter in the presence of hydrogenous materials such as coal, crude oil and other hydrocarbon materials depends strongly on the number of hydrogen nuclei present. Consequently, the facility has been adapted for quantification of crude oil in soil contaminated in this work. The description of the facility and analytical procedures for quantification of oil spill in soil contaminated with different amount of crude oil are provided

  4. An investigation of tungsten by neutron activation techniques

    International Nuclear Information System (INIS)

    Svetsreni, R.

    1978-01-01

    This investigation used neutron from Plutonium-Beryllium source (5 curie) to analyse the amount of tungsten in tungsten oxide which was extracted from tungsten ores, slag and tungsten alloy of tungsten iron and carbon. The technique of neutron activation analysis with NaI(Tl) gamma detector 3'' x 3'' and 1024 multichannel analyzer. The dilution technique was used by mixing Fe 2 O 3 or pure sand into the sample before irradiation. In this study self shielding effect in the analysis of tungsten was solved and the detection limit of the tungsten in the sample was about 0.5%

  5. Neutron radiography activity in the european program cost 524: Neutron imaging techniques

    International Nuclear Information System (INIS)

    Chirco, P.; Bach, P.; Lehmann, E.; Balasko, M.

    2001-01-01

    COST is a framework for scientific and technical cooperation, allowing the coordination of national research on a European level, including 32 member countries. Participation of institutes from non-COST countries is possible. From an initial 7 Actions in 1971, COST has grown to 200 Actions at the beginning of 2000. COST Action 524 is under materials domain, the title of which being 'Neutron Imaging Techniques for the Detection of Defects in Materials', under the Chairmanship of Dr. P. Chirco (I.N.F.N.). The following countries are represented in the Management Committee of Action 524: Italy, France, Austria, Germany, United Kingdom, Hungary, Switzerland, Spain, Czech Republic, Slovenia, and Russia. The six working groups of this Action are working respectively on standardization of neutron radiography techniques, on aerospace application, on civil engineering applications, on comparison and integration of neutron imaging techniques with other NDT, on neutron tomography, and on non radiographic techniques such as neutron scattering techniques. A specific effort is devoted to standardization issues, with respect to other non European standards. Results of work performed in the COST frame are published or will be published in the review INSIGHT, edited by the British Institute of Non Destructive Testing

  6. Neutron thermalization in absorbing infinite homogeneous media: theoretical methods; Methodes theoriques pour l'etude de la thermalisation des neutrons dans les milieux absorbants infinis et homogenes

    Energy Technology Data Exchange (ETDEWEB)

    Cadilhac, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-11-15

    After a general survey of the theory of neutron thermalization in homogeneous media, one introduces, through a proper formulation, a simplified model generalizing both the Horowitz model (generalized heavy free gas approximation) and the proton gas model. When this model is used, the calculation of spectra is reduced to the solution of linear second order differential equations. Since it depends on two arbitrary functions, the model gives a good approximation of any usual moderator for reactor physics purposes. The choice of these functions is discussed from a theoretical point of view; a method based on the consideration of the first two moments of the scattering law is investigated. Finally, the possibility of discriminating models by using experimental informations is considered. (author) [French] Apres un passage en revue de generalites sur la thermalisation des neutrons dans les milieux homogenes, on developpe un formalisme permettant de definir et d'etudier un modele simplifie de thermaliseur. Ce modele generalise l'approximation proposee par J. HOROWITZ (''gaz lourd generalise'') et comporte comme cas particulier le modele ''hydrogene gazeux monoatomique''. Il ramene le calcul des spectres a la resolution d'equations differentielles lineaires du second ordre. Il fait intervenir deux fonctions arbitraires, ce qui lui permet de representer les thermaliseurs usuels de facon satisfaisante pour les besoins de la physique des reacteurs. L'ajustement theorique de ces fonctions est discute; on etudie une methode basee sur la consideration des deux premiers moments de la loi de diffusion. On envisage enfin la possibilite de discriminer les modeles d'apres des renseignements d'origine experimentale. (auteur)

  7. Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

    International Nuclear Information System (INIS)

    Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

    2010-01-01

    These days, the demand to use nuclear resources efficiently is growing for long-term energy supply and also for solving the green house problem. It is indispensable to develop technologies to reduce environmental load with the nuclear energy supply for sustainable development of human beings. In this regard, the development of the fast breeder reactor (FBR) is preferable to utilize nuclear resources effectively and also to burn minor actinides which possess very long toxicity for more than thousands years if they are not extinguished. As one of the FBR developing works in Japan this phase I study started in 2006 to introduce hafnium (Hf) hydride and Gadolinium-Zirconium (Gd-Zr) hydride as new control materials in FBR. By adopting them, the FBR core control technology is improved by two ways. One is extension of control rod life time by using long life Hf hydride which leads to reduce the fabrication and disposal cost and the other is reduction of the excess reactivity by adopting Gd-Zr hydride which leads to reduce the number of control rods and simplifies the core upper structure. This three year study was successfully completed and the following results were obtained. The core design was performed to examine the applicability of the Hf hydride absorber to Japanese Sodium Fast Reactor (JSFR) and it is clarified that the control rod life time can be prolonged to 6 years by adopting Hf hydride and the excess reactivity of the beginning of the core cycle can be reduced to half and the number of the control rods is also reduced to half by using the Gd-Zr hydride burnable poison. The safety analyses also certified that the core safety can be maintained with the same reliability of JSFR Hf hydride and Gd-Zr hydride pellets were fabricated in good manner and their basic features for design use were measured by using the latest devices such as SEM-EDX. In order to reduce the hydrogen transfer through the stainless steel cladding a new technique which shares calorizing

  8. Technique-dependent decrease in thyroid absorbed dose for dental radiography

    International Nuclear Information System (INIS)

    Wood, R.E.; Bristow, R.G.; Clark, G.M.; Nussbaum, C.; Taylor, K.W.

    1989-01-01

    A LiF thermoluminescent dosimetry (TLD) system, calibrated in the tissue of interest with the beam used for experimentation, was employed to investigate dosages (muGy) to the thyroid region of an anthropomorphic phantom resultant from two dental complete-mouth radiographic procedures. Both techniques were compared in terms of dosages associated with combinations of lead apron and thyroid collar shielding while using a 70-kVp or 90-kVp x-ray beam for a 20-film complete-mouth series. Lead shielding significantly decreased the dose to the thyroid using both techniques (p less than 0.05). The use of the 90-kVp beam resulted in a significant reduction in the thyroid absorbed dose when using the bisecting angle technique (p less than 0.05) but caused a significant increase in the thyroid absorbed dose when the paralleling technique was used (p less than 0.05). The implementation of higher kilovoltage techniques in dental offices must therefore be dependent on the radiographic technique employed

  9. Measurement of the neutron detection efficiency of a 80% absorber-20% scintillating fibers calorimeter

    Energy Technology Data Exchange (ETDEWEB)

    Anelli, M.; Bertolucci, S. [Laboratori Nazionali di Frascati dell' INFN, Via E.Fermi 40, I-00044 Frascati (Italy); Bini, C., E-mail: cesare.bini@roma1.infn.i [Dipartimento di Fisica, Sapienza Universita di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); INFN Sezione di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); Branchini, P. [INFN Sezione di Roma Tre, Via della Vasca Navale, 84 I-00146 Roma (Italy); Corradi, G.; Curceanu, C. [Laboratori Nazionali di Frascati dell' INFN, Via E.Fermi 40, I-00044 Frascati (Italy); De Zorzi, G.; Di Domenico, A. [Dipartimento di Fisica, Sapienza Universita di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); INFN Sezione di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); Di Micco, B. [Dipartimento di Fisica dell' Universita ' Roma Tre' , Via della Vasca Navale, 84 I-00146 Roma (Italy); INFN Sezione di Roma Tre, Via della Vasca Navale, 84 I-00146 Roma (Italy); Ferrari, A. [Institute of Safety Research and Institute of Radiation Physics, Forschungszentrum Dresden-Rossendorf, PF 510119, 01314 Dresden (Germany); Fiore, S. [Dipartimento di Fisica, Sapienza Universita di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); INFN Sezione di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); Gauzzi, P., E-mail: paolo.gauzzi@roma1.infn.i [Dipartimento di Fisica, Sapienza Universita di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); INFN Sezione di Roma, P.le A.Moro, 2 I-00185 Roma (Italy); Giovannella, S.; Happacher, F. [Laboratori Nazionali di Frascati dell' INFN, Via E.Fermi 40, I-00044 Frascati (Italy); Iliescu, M. [Laboratori Nazionali di Frascati dell' INFN, Via E.Fermi 40, I-00044 Frascati (Italy); ' Horia Hulubei' National Institute of Physics and Nuclear Engineering, Str. Atomistilor no. 407, P.O. Box MG-6 Bucharest-Magurele (Romania); Luca, A.; Martini, M.; Miscetti, S. [Laboratori Nazionali di Frascati dell' INFN, Via E.Fermi 40, I-00044 Frascati (Italy)

    2011-01-21

    The neutron detection efficiency of a sampling calorimeter made of 1 mm diameter scintillating fibers embedded in a lead/bismuth structure has been measured at the neutron beam of The Svedberg Laboratory at Uppsala. A significant enhancement of the detection efficiency with respect to a bulk organic scintillator detector with the same thickness is observed.

  10. Neutron spectral modulation as a new thermal neutron scattering technique. Pt. 1

    International Nuclear Information System (INIS)

    Ito, Y.; Nishi, M.; Motoya, K.

    1982-01-01

    A thermal neutron scattering technique is presented based on a new idea of labelling each neutron in its spectral position as well as in time through the scattering process. The method makes possible the simultaneous determination of both the accurate dispersion relation and its broadening by utilizing the resolution cancellation property of zero-crossing points in the cross-correlated time spectrum together with the Fourier transform scheme of the neutron spin echo without resorting to the echoing. The channel Fourier transform applied to the present method also makes possible the determination of the accurate direct energy scan profile of the scattering function with a rather broad incident neutron wavelength distribution. Therefore the intensity sacrifice for attaining high accurarcy is minimized. The technique is used with either a polarized or unpolarized beam at the sample position with no precautions against beam depolarization at the sample for the latter case. Relative time accurarcy of the order of 10 -3 to 10 -4 may be obtained for the general dispersion relation and for the quasi-elastic energy transfers using correspondingly the relative incident neutron wavelength spread of 10 to 1% around an incident neutron energy of a few meV. (orig.)

  11. Neutron diffraction technique as a method for material studies

    International Nuclear Information System (INIS)

    Belhorma, B.; Labrim, H.; Gandou, Z.

    2010-01-01

    The Morocco's first Nuclear Research has been constructed in CNESTEN. The reactor divergence has been tested, and the nominal power of 2MW was successfully achieved. The reactor has 4 beam ports two of them are projected for neutron scattering. Such technique allows studying the crystallographic and magnetic structures of materials using the thermal neutrons produced in the reactor. the powder diffractometer has been designed. Component reception and installation procedures are in progress. The second experiment consists on small angle neutron scattering that allows the study of soft matter and polymers in the range of 1-50 nm. The third technique that can complete the two previous is the 4-circle neutron spectrometry which is designed mainly to study structural properties of the mono-crystalline material and texture.This technique is complementary to the X-ray diffraction already available in CNESTEN. Some applications of this technique are: --to determine the crystallographic and magnetic structure of polycrystalline materials.-- to study the texture in metals and alloys.-- to perform holography measurement

  12. RBE/absorbed dose relationship of d(50)-Be neutrons determined for early intestinal tolerance in mice

    International Nuclear Information System (INIS)

    Gueulette, J.; Wambersie, A.

    1978-01-01

    RBE/absorbed dose relationship of d(50)-Be neutrons (ref.: 60 Co) was determined using intestinal tolerance in mice (LD50) after single and fractionated irradiation. RBE is 1.8 for a single fraction (about 1000 rad 60 Co dose); it increases when decreasing dose and reaches the plateau value of 2.8 for a 60 Co dose of about 200 rad. This RBE value is used for the clinical applications with the cyclotron 'Cyclone' at Louvain-la-Neuve [fr

  13. Spectra and absorbed dose by photo-neutrons in a solid water mannequin exposed to a Linac of 15 MV

    International Nuclear Information System (INIS)

    Benites R, J.; Vega C, H. R.; Velazquez F, J.

    2012-10-01

    Using Monte Carlo methods was modeled a solid water mannequin; according to the ICRU 44 (1989), Tissue substitutes in radiation dosimetry and measurements, of the International Commission on Radiation Units and Measurements; Report 44. This material Wt 1 is made of H (8.1%), C (67.2%), N (2.4%), O (19.9%), Cl (0.1%), Ca (2.3%) and its density is of 1.02 gr/cm 3 . The mannequin was put instead of the patient, inside the treatment room and the spectra and absorbed dose were determined by photo-neutrons exposed to a Linac of 15 MV. (Author)

  14. Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

    International Nuclear Information System (INIS)

    Overbury, Steven H.; Coates, Leighton; Herwig, Kenneth W.; Kidder, Michelle

    2011-01-01

    This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

  15. Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

    Energy Technology Data Exchange (ETDEWEB)

    Overbury, Steven {Steve} H [ORNL; Coates, Leighton [ORNL; Herwig, Kenneth W [ORNL; Kidder, Michelle [ORNL

    2011-10-01

    This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

  16. Influence of the neutron flux shape on the value of absorbed neutron dose; Uticaj oblika neutronskog spektra na vrednost apsorbovane doze neutrona

    Energy Technology Data Exchange (ETDEWEB)

    Miric, I; Miric, P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    This paper deals with the study od specific doses dependence on the type and approximation procedures of neutron spectra. Values of specific dose rates (dose per neutron cm{sub 2}) were analysed for neutron spectra from RB reactor in Vinca, Crac facility in Valduc (France) and HPRR reactor in Oak Ridge (USA). Data used in this analysis were obtained by methods used in Harwell (AERE), Oak Ridge (ORNL), Chalk River (AECL), CEN de Cadarache (CEA) and in the Boris Kidric Institute (IBK). Specific absorbed neutron doses were determined for each of the estimated spectra and presented in the form of kerma/(n.cm{sup -2}) and rad/((n.cm{sup -2}) units. The obtained results have shown the influence of the flux approximation procedure on the values of conversion factors for obtaining neutron doses from neutron flux. U okviru ovog rada radjeno je na ispitivanju zavisnosti specificnih doza od vrste i nacina aproksimacije neutronskog spektra. U radu su analizirane vrednosti specificnih doza (doza po n.cm{sup -2}) za neutronske spektre koji se dobijaju oko sledecih nuklearnih postrojenja: reaktora RB u Vinci, postrojenja CRAC u Valduc-u (Francuska), reaktora HPRR u Oak Ridge-u (SAD). Za analizu su korisceni podaci dobijeni metodama koje se koriste u nuklearnim centrima Harwell (AERE), Oak Ridge-u (ORNL), Chalk River-u (AECL), CEN de Cadarache (CEA) i Institutu Boris Kidric (IBK). Za svaki procenjeni spektar odredjene su specificne apsorbovane doze neutrona izrazene u kerma/(n.cm{sup -2}) i rad/(n.cm{sup -2}) jedinicama. Dobijeni rezultati su pokazali koliko nacin aproksimacije spektra utice na vrednost konverzionih faktora koji sluze za prelazak sa fluksa na dozu neutrona (author)

  17. Alternative technique to neutron probe calibration in situ

    International Nuclear Information System (INIS)

    Encarnacao, F.; Carneiro, C.; Dall'Olio, A.

    1990-01-01

    An alternative technique of neutron probe calibration in situ was applied for Podzolic soil. Under field condition, the neutron probe calibration was performed using a special arrangement that prevented the lateral movement of water around the access tube of the neutron probe. During the experiments, successive amounts of water were uniformly infiltrated through the soil profile. Two plots were set to study the effect of the plot dimension on the slope of the calibration curve. The results obtained shown that the amounts of water transferred to the soil profile were significantly correlated to the integrals of count ratio along the soil profile on both plots. In consequence, the slope of calibration curve in field condition was determined. (author)

  18. Active neutron technique for detecting attempted special nuclear material diversion

    International Nuclear Information System (INIS)

    Smith, G.W.; Rice, L.G. III.

    1979-01-01

    The identification of special nuclear material (SNM) diversion is necessary if SNM inventory control is to be maintained at nuclear facilities. (Special nuclear materials are defined for this purpose as either 235 U of 239 Pu.) Direct SNM identification by the detection of natural decay or fission radiation is inadequate if the SNM is concealed by appropriate shielding. The active neutron interrogation technique described combines direct SNM identification by delayed fission neutron (DFN) detection with implied SNM detection by the identification of materials capable of shielding SNM from direct detection. This technique is being developed for application in an unattended material/equipment portal through which items such as electronic instruments, packages, tool boxes, etc., will pass. The volume of this portal will be 41-cm wide, 53-cm high and 76-cm deep. The objective of this technique is to identify an attempted diversion of at least 20 grams of SNM with a measurement time of 30 seconds

  19. Neutron Diffusion in a Space Lattice of Fissionable and Absorbing Materials

    Science.gov (United States)

    Feynman, R. P.; Welton, T. A.

    1946-08-27

    Methods are developed for estimating the effect on a critical assembly of fabricating it as a lattice rather than in the more simply interpreted homogeneous manner. An idealized case is discussed supposing an infinite medium in which fission, elastic scattering and absorption can occur, neutrons of only one velocity present, and the neutron m.f.p. independent of position and equal to unity with the unit of length used.

  20. Characterization of European sword blades through neutron imaging techniques

    Science.gov (United States)

    Salvemini, F.; Grazzi, F.; Peetermans, S.; Gener, M.; Lehmann, E. H.; Zoppi, M.

    2014-09-01

    In the present work, we have studied two European rapier blades, dating back to the period ranging from the Late Renaissance to the Early Modern Age (about 17th to 18th century). In order to determine variation in quality and differences in technology, a study was undertaken with the purpose to observe variations in the blade microstructure (and consequently in the construction processes). The samples, which in the present case were expendable, have been investigated, preliminarily, through standard metallography and then by means of white beam and energy-selective neutron imaging. The comparison of the results, using the two techniques, turned out to be satisfactory, with a substantial quantitative agreement of the results obtained with the two techniques, and show the complementarity of the two methods. Metallography has been considered up to now the method of choice for metal material characterization. The correspondence between the two methods, as well as the non-invasive character of the neutron-based techniques and its possibility to obtain 3D reconstruction, candidate neutron imaging as an important and quantitatively reliable technique for metal characterization.

  1. Studies of molecular dynamics with neutron scattering techniques. Part of a coordinated programme on neutron scattering techniques

    International Nuclear Information System (INIS)

    Vinhas, L.A.

    1980-05-01

    Molecular dynamics was studied in samples of tert-butanol, cyclohexanol and methanol, using neutron inelastic and quasi-elastic techniques. The frequency spectra of cyclohexanol in crystalline phase were interpreted by assigning individual energy peaks to hindered rotation of molecules, lattice vibration, hydrogen bond stretching and ring bending modes. Neutron quasi-elastic scattering measurements permitted the testing of models for molecular diffusion as a function of temperature. The interpretation of neutron incoherent inelastic scattering on methanol indicated the different modes of molecular dynamics in this material; individual inelastic peaks in the spectra could be assigned to vibrations of crystalline lattice, stretching of hydrogen bond and vibrational and torsional modes of CH 3 OH molecule. The results of the experimental work on tertbutanol indicate two distinct modes of motion in this material: individual molecular librations are superposed to a cooperative rotation diffusion which occurs both in solid and in liquid state

  2. A study on artificial rare earth (RE2O3) based neutron absorber.

    Science.gov (United States)

    Kim, Kyung-O; Kyung Kim, Jong

    2015-11-01

    A new concept of a neutron absorption material (i.e., an artificial rare earth compound) was introduced for criticality control in a spent fuel storage system. In particular, spent nuclear fuels were considered as a potential source of rare earth elements because the nuclear fission of uranium produces a full range of nuclides. It was also found that an artificial rare earth compound (RE2O3) as a High-Level Waste (HLW) was naturally extracted from pyroprocessing technology developed for recovering uranium and transuranic elements (TRU) from spent fuels. In this study, various characteristics (e.g., activity, neutron absorption cross-section) were analyzed for validating the application possibility of this waste compound as a neutron absorption material. As a result, the artificial rare earth compound had a higher neutron absorption probability in the entire energy range, and it can be used for maintaining sub-criticality for more than 40 years on the basis of the neutron absorption capability of Boral™. Therefore, this approach is expected to vastly improve the efficiency of radioactive waste management by simultaneously keeping HLW and spent nuclear fuel in a restricted space. Copyright © 2015 Elsevier Ltd. All rights reserved.

  3. A study on artificial rare earth (RE2O3) based neutron absorber

    International Nuclear Information System (INIS)

    KIM, Kyung-O; Kyung KIM, Jong

    2015-01-01

    A new concept of a neutron absorption material (i.e., an artificial rare earth compound) was introduced for criticality control in a spent fuel storage system. In particular, spent nuclear fuels were considered as a potential source of rare earth elements because the nuclear fission of uranium produces a full range of nuclides. It was also found that an artificial rare earth compound (RE 2 O 3 ) as a High-Level Waste (HLW) was naturally extracted from pyroprocessing technology developed for recovering uranium and transuranic elements (TRU) from spent fuels. In this study, various characteristics (e.g., activity, neutron absorption cross-section) were analyzed for validating the application possibility of this waste compound as a neutron absorption material. As a result, the artificial rare earth compound had a higher neutron absorption probability in the entire energy range, and it can be used for maintaining sub-criticality for more than 40 years on the basis of the neutron absorption capability of Boral™. Therefore, this approach is expected to vastly improve the efficiency of radioactive waste management by simultaneously keeping HLW and spent nuclear fuel in a restricted space. - Highlights: • Quantitative analysis of rare earth elements in PWR spent fuels. • Extraction of artificial rare earth compound using pyroprocessing technology. • Characteristic analysis of artificial rare earth elements. • Performance evaluation of artificial rare earth for criticality control.

  4. Dynamic simulation of a two-phase control absorber for neutron flux regulation in a nuclear reactor

    International Nuclear Information System (INIS)

    Plourde, J.A.; Lepp, R.M.

    1979-08-01

    A dynamic simulation of the two-phase control absorber being proposed for future Canadian nuclear power reactors has been developed at Chalk River Nuclear Laboratories. The model, implemented on a hybrid computer, was developed to study absorber dynamics at different circuit operating conditions and with different circuit configurations. The simulation is modular, with as much correspondence as possible between individual modules and the physical entities. The dynamics of several of the modules are described by partial differential equations, with space and time as independent variables. These are solved via the Continuous Space/Discrete Time technique. The simulation has been validated with data from the Two-Phase Absorber Experimental (TOPAX) Rig installed at the ZED-2 test reactor. (author)

  5. Development of active neutron interrogation techniques at Harwell

    International Nuclear Information System (INIS)

    Armitage, B.H.; Chard, P.M.J.; Packer, T.W.; Swinhoe, M.T.; Syme, D.B.

    1990-01-01

    Active neutron interrogation techniques capable of measuring the fissile content of a range of waste drum sizes and contents have been developed at Harwell. This paper describes measurements which have been made to investigate the behaviour of these assay systems for the difficult case of concreted waste in a heterogeneous matrix. The drums have been measured using a Cf shuffler and a differential die-away system, with supporting information obtained from a segmented gamma-scanner. Good correspondence has been observed between the two different neutron interrogation techniques. It was concluded that the measurement of highly heterogeneous wastes is likely to be more effective if calibration can be undertaken with representative artificial matrices. Further measurement and analysis remains to be undertaken

  6. Study of corrosion in aluminium using neutron radiography technique

    International Nuclear Information System (INIS)

    Islam, M.N.; Alam, M.K.; Saklayen, M.A.; Ahsan, M.H.; Islam, S.M.A.; Zaman, M.A.

    2000-01-01

    Neutron radiography technique has been adopted for detection of corrosion in aluminium by filling artificially made holes on aluminium slab with Al(OH) 3 . The contrast between the optical densities of corrosion products and aluminium slab was assessed from the densitometric measurements. Variation of optical density difference with sample thickness has also been studied. The results confirm that approximately 0.039 mm thick corrosion products having diameter 10 mm can easily be detected in 2 cm thick aluminium slab. The linear attenuation coefficient of Al(OH) 3 has been obtained as 0.9447. From the present investigation it is confirmed that film neutron radiography (NR) technique is helpful for investigation of Al(OH) 3 type corrosion product in aluminium. (author)

  7. 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

  8. Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

    International Nuclear Information System (INIS)

    Aizawa, O.

    1993-01-01

    In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

  9. Experiments at the GELINA facility for the validation of the self-indication neutron resonance densitometry technique

    Directory of Open Access Journals (Sweden)

    Rossa Riccardo

    2017-01-01

    Full Text Available Self-Indication Neutron Resonance Densitometry (SINRD is a passive non-destructive method that is being investigated to quantify the 239Pu content in a spent fuel assembly. The technique relies on the energy dependence of total cross sections for neutron induced reaction. The cross sections show resonance structures that can be used to quantify the presence of materials in objects, e.g. the total cross-section of 239Pu shows a strong resonance close to 0.3 eV. This resonance will cause a reduction of the number of neutrons emitted from spent fuel when 239Pu is present. Hence such a reduction can be used to quantify the amount of 239Pu present in the fuel. A neutron detector with a high sensitivity to neutrons in this energy region is used to enhance the sensitivity to 239Pu. This principle is similar to self-indication cross section measurements. An appropriate detector can be realized by surrounding a 239Pu-loaded fission chamber with appropriate neutron absorbing material. In this contribution experiments performed at the GELINA time-of-flight facility of the JRC at Geel (Belgium to validate the simulations are discussed. The results confirm that the strongest sensitivity to the target material was achieved with the self-indication technique, highlighting the importance of using a 239Pu fission chamber for the SINRD measurements.

  10. NET European Network on Neutron Techniques Standardization for Structural Integrity

    International Nuclear Information System (INIS)

    Youtsos, A.

    2004-01-01

    Improved performance and safety of European energy production systems is essential for providing safe, clean and inexpensive electricity to the citizens of the enlarged EU. The state of the art in assessing internal stresses, micro-structure and defects in welded nuclear components -as well as their evolution due to complex thermo-mechanical loads and irradiation exposure -needs to be improved before relevant structural integrity assessment code requirements can safely become less conservative. This is valid for both experimental characterization techniques and predictive numerical algorithms. In the course of the last two decades neutron methods have proven to be excellent means for providing valuable information required in structural integrity assessment of advanced engineering applications. However, the European industry is hampered from broadly using neutron research due to lack of harmonised and standardized testing methods. 35 European major industrial and research/academic organizations have joined forces, under JRC coordination, to launch the NET European Network on Neutron Techniques Standardization for Structural Integrity in May 2002. The NET collaborative research initiative aims at further development and harmonisation of neutron scattering methods, in support of structural integrity assessment. This is pursued through a number of testing round robin campaigns on neutron diffraction and small angle neutron scattering - SANS and supported by data provided by other more conventional destructive and non-destructive methods, such as X-ray diffraction and deep and surface hole drilling. NET also strives to develop more reliable and harmonized simulation procedures for the prediction of residual stress and damage in steel welded power plant components. This is pursued through a number of computational round robin campaigns based on advanced FEM techniques, and on reliable data obtained by such novel and harmonized experimental methods. The final goal of

  11. Using Neutron-based techniques to investigate battery behaviour

    International Nuclear Information System (INIS)

    Pramudita, James C.; Goonetilleke, Damien; Sharma, Neeraj; Peterson, Vanessa K.

    2016-01-01

    The extensive use of portable electronic devices has given rise to increasing demand for reliable high energy density storage in the form of batteries. Today, lithium-ion batteries (LIBs) are the leading technology as they offer high energy density and relatively long lifetimes. Despite their widespread adoption, Li-ion batteries still suffer from significant degradation in their performance over time. The most obvious degradation in lithium-ion battery performance is capacity fade – where the capacity of the battery reduces after extended cycling. This talk will focus on how in situ time-resolved neutron powder diffraction (NPD) can be used to gain a better understanding of the structural changes which contribute to the observed capacity fade. The commercial batteries studied each feature different electrochemical and storage histories that are precisely known, allowing us to elucidate the tell-tale signs of battery degradation using NPD and relate these to battery history. Moreover, this talk will also showcase the diverse use of other neutron-based techniques such as neutron imaging to study electrolyte concentrations in lead-acid batteries, and the use of quasi-elastic neutron scattering to study Na-ion dynamics in sodium-ion batteries.

  12. In vivo measurements of nitrogen using a neutron activation technique

    International Nuclear Information System (INIS)

    Larsson, L.; Alpsten, M.; Toelli, J.; Drugge, N.; Mattsson, S.

    1986-01-01

    Knowledge of body composition is essential for understanding of many diseases such as obesity, anorexia, cancer, kidney and heart diseases. For many years, total body potassium (TBK) has been used as an estimate of the intracellular protein. In some diseases intracellular- and extracellular protein may vary significantly. Together with TBK, total body nitrogen (TBN) should in these cases be measured to estimate the total protein content. The nitrogen content can be measured by in vivo neutron activation. In this work the authors have used the prompt gamma technique: Thermalized neutrons from a Cf-252-source are captured in (n, δ)-reactions. Prompt 10.8 MeV photons are emitted and can be detected during irradiation. The source is contained in a polyethylene block which forms a collimator surrounded by a phi 1.40 m x 0.80 m water tank. The patient is irradiated from below by a 15 cm x 50 cm neutron field. It is possible to scan the whole patient or to measure a part of the body. A phi 15 cm x 15 cm NaI(T1)-detector is used for detection of the 10.8 MeV photons. The detector is mounted above the patient outside the neutron field

  13. Image reconstruction technique using projection data from neutron tomography system

    Directory of Open Access Journals (Sweden)

    Waleed Abd el Bar

    2015-12-01

    Full Text Available Neutron tomography is a very powerful technique for nondestructive evaluation of heavy industrial components as well as for soft hydrogenous materials enclosed in heavy metals which are usually difficult to image using X-rays. Due to the properties of the image acquisition system, the projection images are distorted by several artifacts, and these reduce the quality of the reconstruction. In order to eliminate these harmful effects the projection images should be corrected before reconstruction. This paper gives a description of a filter back projection (FBP technique, which is used for reconstruction of projected data obtained from transmission measurements by neutron tomography system We demonstrated the use of spatial Discrete Fourier Transform (DFT and the 2D Inverse DFT in the formulation of the method, and outlined the theory of reconstruction of a 2D neutron image from a sequence of 1D projections taken at different angles between 0 and π in MATLAB environment. Projections are generated by applying the Radon transform to the original image at different angles.

  14. Effect of absorber rods on the space-energy distribution of thermal neutrons in water

    International Nuclear Information System (INIS)

    Hussein, A.Z.; Eid, Y.; Hamouda, I.

    1975-01-01

    Thermal neutron spectra have been measured in a vectorial direction with respect to cadmium, boron-filled and copper rod elements. The rods are infinite cylinders, of 21 mm diameter, each separately immersed in an infinite water moderator fed with neutrons from the ET-RR-1 research reactor. Measurement of spectra has been carried out, in the vicinity of the rod elements, at several distances by the time-of-flight method using a chopper and also by intergral flux activation method. The measured spectra near the copper rod were compared with transport calculations of the position-dependent spectrum. The calculations, based on a realistic kernel for water, were found to yield reasonable agreement with experiment. (orig.) [de

  15. Neutron Filter Technique and its use for Fundamental and applied Investigations

    International Nuclear Information System (INIS)

    Gritzay, V.; Kolotyi, V.

    2008-01-01

    At Kyiv Research Reactor (KRR) the neutron filtered beam technique is used for more than 30 years and its development continues, the new and updated facilities for neutron cross section measurements provide the receipt of neutron cross sections with rather high accuracy: total neutron cross sections with accuracy 1% and better, neutron scattering cross sections with 3-6% accuracy. The main purpose of this paper is presentation of the neutron measurement techniques, developed at KRR, and demonstration some experimental results, obtained using these techniques

  16. Synthesis and characterization of super absorbent poly (acrylamide-co-potassium acrylate) hydrogels by radiation technique

    International Nuclear Information System (INIS)

    Erizal

    2010-01-01

    A series of super absorbent hydrogels were prepared from acrylamide (AAm) and potassium acrylate (KA) by gamma irradiation technique at room temperature. The solution containing potassium acrylate 15% and different concentrations of AAm (10-16%) were irradiated by gamma rays (20-40 kGy). The hydrogels produced by irradiation were characterized by fourier transform infra red spectroscopy (FT-IR). The gel fraction, kinetics of swelling and the equilibrium degree of swelling (EDS) were studied. Under irradiation dose of 20 kGy and concentration of AAM 10 %), poly(AAm-co-KA) hydrogel with high gel fraction (99.08%) and very high EDS (420 g/g) were obtained. The capacity of hydrogel to adsorb metal ion Cu 2+ and Fe 3+ were investigated. It is shown than 10 minutes the hydrogel could adsorb Cu 2+ ion up to 95 %, and Fe 3+ ion up to 55 % in 80 minutes. This hydrogel has a potential to be used for soil conditioning and ion metal absorbent. (author)

  17. Two-phase flow measurement by pulsed neutron activation techniques

    International Nuclear Information System (INIS)

    Kehler, P.

    1978-01-01

    The Pulsed Neutron Activation (PNA) technique for measuring the mass flow velocity and the average density of two-phase mixtures is described. PNA equipment can be easily installed at different loops, and PNA techniques are non-intrusive and independent of flow regimes. These features of the PNA technique make it suitable for in-situ measurement of two-phase flows, and for calibration of more conventional two-phase flow measurement devices. Analytic relations governing the various PNA methods are derived. The equipment and procedures used in the first air-water flow measurement by PNA techniques are discussed, and recommendations are made for improvement of future tests. In the present test, the mass flow velocity was determined with an accuracy of 2 percent, and average densities were measured down to 0.08 g/cm 3 with an accuracy of 0.04 g/cm 3 . Both the accuracy of the mass flow velocity measurement and the lower limit of the density measurement are functions of the injected activity and of the total number of counts. By using a stronger neutron source and a larger number of detectors, the measurable density can be decreased by a factor of 12 to .007 g/cm 3 for 12.5 cm pipes, and to even lower ranges for larger pipes

  18. Understanding and predicting the behaviour of silver base neutron absorbers under irradiations; Comprehension et prediction du comportement sous irradiation neutronique d`alliages absorbants a base d`argent

    Energy Technology Data Exchange (ETDEWEB)

    Desgranges, C

    1998-12-31

    The effect of neutron irradiation induced transmutations on the swelling of AgInCd (AIC) alloys used as neutron absorber in the control rods of Pressurized Water Reactors has been studied both experimentally and theoretically. Effective atomic volumes have been determined in synthetic AgCdInSn alloys with various compositions and containing fcc and hc phases, representative of irradiated AIC (Sn is a transmutation product). Swelling is shown to result first from the transmutation of Ag into Cd and of In into Sn, both with larger effective volume than the mother atom, and second from grain boundaries precipitation of s still less dense hc phase when solid solubility of transmuted products is exceeded. For both fcc and hc phases, we have determined profiles at the temperatures in the vicinity of the operating temperature. Unusual characteristics of second phase growth at grain boundaries induced by transmutations are identified on a simple binary alloy model: kinetics is controlled by irradiation temperature which scales diffusivities and flux which scales transmutation rates, as well as by the grain size in the underlying matrix. To address the AgInCdSn alloys, a novel technique is proposed to model diffusion in multicomponent alloys. It is based on a linearization of a simple atomistic model. With a single set of parameters, for each phase, our model well reproduces our interdiffusion measurements in quaternary alloys as well as existing interdiffusion experiments in binary alloys. Finally this diffusion model implemented with a moving interface algorithm is used to model the growth of the second phase induced by transmutation in the AIC under irradiation. (authors) 74 refs.

  19. New laser technique revives old ideas for thermoluminescence neutron dosimetry

    International Nuclear Information System (INIS)

    Braeunlich, P.; Brown, M.; Gasiot, J.; Fillard, J.P.

    1982-01-01

    Laser heating is discussed as a means to evaluate thermoluminescence dosimeters in neutron dosimetry. Direct energy coupling from the photon beam to the phonons of the TL material permits heating of thin layers with rates of temperature increase exceeding 10 4 Ks - 1 . Rapid TLD evaluation will allow the design of dosimetry badges containing a number of different small thin film TLD elements in various orientations and behind appropriate filters, hydrogenous radiators, etc. Desired redundance is readily possible by using back-up TLDs for every specific task. Reading occurs with a scanning laser beam rather than by mechanically manipulating the TLD toward a fixed heat source. Improvements in the signal-to-noise ratio of up to a factor of 1000 are readily obtained. Thus, sensitive thin-film TLDs can be designed with negligible self-shielding for thermal neutrons in albedo applications and with known, nearly energy dependent cavity correction factors for dosimetry in mixed n-#betta# fields. Due to the greatly increased sensitivity possible with fast laser heating, significant advances are expected in the fast neutron dosimetry techniques which are based on hydrogeneous proton radiators or LET-dependent slow peak formation

  20. Error reduction techniques for Monte Carlo neutron transport calculations

    International Nuclear Information System (INIS)

    Ju, J.H.W.

    1981-01-01

    Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

  1. Measurement of the neutron capture cross-section of 232Th using the neutron activation technique

    International Nuclear Information System (INIS)

    Naik, H.; Singh, Sarbjit; Goswami, A.; Manchanda, V.K.; Prajapati, P.M.; Surayanarayana, S.V.; Nayak, B.K.; Sharma, S.C.; Jagadeesan, K.C.; Thakare, S.V.; Raj, D.; Ganesan, S.; Mulik, V.K.; Sivashankar, B.S.; Mukherjee, S.

    2011-01-01

    The 232 Th(n, γ) reaction cross-section at average neutron energies of 3.7±0.3 MeV and 9.85±0.38 MeV from the 7 Li(p, n) reaction has been determined for the first time using activation and off-line γ -ray spectrometric technique. The 232 Th(n, 2n) reaction cross-section at the average neutron energy of 9.85±0.38 MeV has been also determined using the same technique. The experimentally determined 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were compared with the evaluated data of ENDF/B-VII, JENDL-4.0 and JEFF-3.1 and were found to be in good agreement. The present data along with literature data in a wide range of neutron energies were interpreted in terms of competition between different reaction channels including fission. The 232 Th(n, γ) and 232 Th(n, 2n) reaction cross-sections were also calculated theoretically using the TALYS 1.2 computer code and were found to be slightly higher than the experimental data. (orig.)

  2. The application of radiotracer technique for preconcentration neutron activation analysis

    International Nuclear Information System (INIS)

    Wang Xiaolin; Chen Yinliang; Sun Ying; Fu Yibei

    1995-01-01

    The application of radiotracer technique for preconcentration neutron activation analysis (Pre-NAA) are studied and the method for determination of chemical yield of Pre-NAA is developed. This method has been applied to determination of gold, iridium and rhenium in steel and rock samples and the contents of noble metal are in the range of 1-20 ng·g -1 (sample). In addition, the accuracy difference caused by determination of chemical yield between RNAA and Pre-NAA are also discussed

  3. Simultaneous and integrated neutron-based techniques for material analysis of a metallic ancient flute

    International Nuclear Information System (INIS)

    Festa, G; Andreani, C; Pietropaolo, A; Grazzi, F; Scherillo, A; Barzagli, E; Sutton, L F; Bognetti, L; Bini, A; Schooneveld, E

    2013-01-01

    A metallic 19th century flute was studied by means of integrated and simultaneous neutron-based techniques: neutron diffraction, neutron radiative capture analysis and neutron radiography. This experiment follows benchmark measurements devoted to assessing the effectiveness of a multitask beamline concept for neutron-based investigation on materials. The aim of this study is to show the potential application of the approach using multiple and integrated neutron-based techniques for musical instruments. Such samples, in the broad scenario of cultural heritage, represent an exciting research field. They may represent an interesting link between different disciplines such as nuclear physics, metallurgy and acoustics. (paper)

  4. Study of neutron absorbing microspheres in research reactors - Metal systems wear

    International Nuclear Information System (INIS)

    Gana Watkins, Ignacio A.; Silin, Nicolas; Prado, Miguel O.; Mazufri, Claudio

    2012-01-01

    Now-a-days, it is increasingly common for nuclear power plants, as well as research reactors, to be designed and built with an alternative safety system aside from control rods. The acids and/or salts in solution injection systems is most frequently used. However, these systems present several implementation and operation problems due to the physical and chemical properties of the used compounds. After analyzing these drawbacks, we developed a new alternative safety system that contains the absorbing element isolated from the aqueous medium. In this context, it's proposed the use of aluminum borosilicate microspheres. The current paper presents erosion wear experiments to determine under which conditions microspheres can be considered as a potential component of a secondary shut down system in a nuclear facility (author))

  5. Integrating techniques for neutron dosimetry in Linac 18 MV; Integrando tecnicas para dosimetria de neutrones en un Linac de 18 MV

    Energy Technology Data Exchange (ETDEWEB)

    Ceron R, P. V.; Diaz G, J. A. I.; Rivera M, T. [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Av. Legaria 694, 11500 Mexico D. F. (Mexico); Paredes G, L. C. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this paper thermoluminescent dosimetry, analytical techniques and Monte Carlo calculations were used to estimate the neutron dose equivalent in a radiotherapy room with a linear electron accelerator of 18 MV. The equivalent dose was measured at isocenter to 1.42 m of target and at the entrance of the labyrinth of the room of a Novalis Tx. The neutron detectors were constructed with pairs of thermoluminescent dosimeters TLD 600 ({sup 6}LiF: Mg, Ti) and TLD 700 ({sup 7}LiF: Mg, Ti) which are placed inside a paraffin sphere of 20 cm in diameter. These measurements enabled the calculation of equivalent dose in the gate and the source term, using the relationships contained in the NCRP-151. Through the models carried out with the code MCNPX the absorbed dose distribution with regard to depth in a paraffin phantom are included and the neutron spectrum produced by the head, taking into account the geometry and component materials. The results are in the order of neutron milli sievert by gray of X-rays (mSv/Gy x) which are in the same order as those found in other reports for different accelerators. (Author)

  6. Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

    International Nuclear Information System (INIS)

    Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

    1984-07-01

    The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

  7. Study on bioavailability of dietary iron of women by using activable isotopic tracer and neutron activation analysis techniques

    International Nuclear Information System (INIS)

    Zhang Yangmei; Ni Bangfa; Tian Weizhi; Wang Pingsheng; Cao Lei

    2002-01-01

    The bioavailability of diet iron of 10 healthy young women in Beijing area is studied by using two enriched isotopes 54 Fe and 58 Fe, and neutron activation analysis techniques. The abundance of 54 Fe and 58 Fe is 61.4% and 23.4%, respectively. In additional, the atomic absorption spectrometry is employed to measure total iron in fecal samples. Dysprosium, rarely absorbed by human body, is used to monitor the residence time of tracer isotopes in order to collect the fecal samples completely. The results show that the bioavailability of dietary iron in young women is (14.9 +- 3.9)%

  8. Measurements of fusion neutron yields by neutron activation technique: Uncertainty due to the uncertainty on activation cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Via E. Fermi, 45, 00044 Frascati, Rome (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Sjöstrand, Henrik; Conroy, Sean [Department of Physics and Astronomy, Uppsala University, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-07-11

    The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields.

  9. Neutron flux characterization of californium-252 Neutron Research Facility at the University of Texas - Pan American by nuclear analytical technique

    Science.gov (United States)

    Wahid, Kareem; Sanchez, Patrick; Hannan, Mohammad

    2014-03-01

    In the field of nuclear science, neutron flux is an intrinsic property of nuclear reaction facilities that is the basis for experimental irradiation calculations and analysis. In the Rio Grande Valley (Texas), the UTPA Neutron Research Facility (NRF) is currently the only neutron facility available for experimental research purposes. The facility is comprised of a 20-microgram californium-252 neutron source surrounded by a shielding cascade containing different irradiation cavities. Thermal and fast neutron flux values for the UTPA NRF have yet to be fully investigated and may be of particular interest to biomedical studies in low neutron dose applications. Though a variety of techniques exist for the characterization of neutron flux, neutron activation analysis (NAA) of metal and nonmetal foils is a commonly utilized experimental method because of its detection sensitivity and availability. The aim of our current investigation is to employ foil activation in the determination of neutron flux values for the UTPA NSRF for further research purposes. Neutron spectrum unfolding of the acquired experimental data via specialized software and subsequent comparison for consistency with computational models lends confidence to the results.

  10. The behavior of moisture content in Durian after harvesting by neutron reflection and transmission techniques

    International Nuclear Information System (INIS)

    Chimoye, T.; Fuangfoong, M.

    1998-01-01

    The study aimed at development of a neutron reflection and transmission technique to determine moisture content in Durian fruit as a function of time after harvesting. A system of a 3 mCi Am-Be neutron source with a BF 3 detector as a neutron probe was developed. The results obtained were validated using weighting method

  11. Radiation shielding material characterization by non-destructive neutron radiography technique

    International Nuclear Information System (INIS)

    Hafizal Yazid; Azali Muhammad; Abdul Aziz Mohamed; Rafhayudi Jamro; Hishamuddin Husain

    2007-01-01

    Shielding property of boronated rubber was characterized easily by the use of neutron radiography technique. For 10 phr of boron carbide in the natural rubber composite, the ability to completely shield against neutron was found to have 8mm thickness and above for the neutron flux of 1.04 x 10 5 n/cm 2 s (author)

  12. Use of the associated particle technique in the fast neutron spectroscopy

    International Nuclear Information System (INIS)

    Aquirre O, G.A.

    1978-01-01

    Selecting a neutrons monoenergetic source it was found that the nuclear reaction D(d,n) 3 He can be used to measure nuclear sections and differentials in elastic nuclear reactions through the associated particle technique; the neutron beam energy is directly determined in time of flight spectrum of the neutron. The flux is determined by the number of 3 He ions observed in the charged particle spectrum. The neutron flux can be increased increasing the solid angle of the neutrons beam in two magnitude orders according to the results of neutrons beam profile measures. (author)

  13. Neutron scattering techniques in the examination of recycled aggregate concrete

    International Nuclear Information System (INIS)

    Krezel, A.; Alabaster, P.; Bakshi, E.; McManus, K.

    1999-01-01

    Full text: Researchers at Swinburne University of Technology (SUT) have undertaken a research project aiming initially at better understanding the effects of any chemical impurities in Recycled Concrete Aggregate (RCA) on the microstructure development of Recycled Aggregate Concrete (RAC). Furthermore, a porosity of RCA and RAC and its effect on the acoustic performance and mechanical properties is being investigated. A number of conventional tests have been employed to examine the porosity of the aggregate and concrete made from RCA ranging from Volume of Permeable Voids test, through nitrogen adsorption to scanning electron microscopy. These tests are performed at SUT to characterise pores structure including pore size and volume as well as their surface area. The preparation of samples differs for the various tests, and this is a main reason contributing to inconsistencies in the results from these tests. None-the-less the results indicate strong positive correlation of inherent and purposely introduced porosity in RAC to its sound absorption capacities. Some inconsistency in the results is also due to the complexity of concrete itself compounded by the use of recycled material. However, the research has been granted a Grant from Australian Institute of Nuclear Science and Engineering (AINSE) which allows to conduct RAC examination using Small Angle Neutron Scattering (SANS). This neutron scattering technique characterises pore structure in a non-destructive manner. The results from this method should augment these obtained from conventional methods

  14. A Comprehensive Review of the Techniques on Regenerative Shock Absorber Systems

    Directory of Open Access Journals (Sweden)

    Ran Zhang

    2018-05-01

    Full Text Available In this paper, the current technologies of the regenerative shock absorber systems have been categorized and evaluated. Three drive modes of the regenerative shock absorber systems, namely the direct drive mode, the indirect drive mode and hybrid drive mode are reviewed for their readiness to be implemented. The damping performances of the three different modes are listed and compared. Electrical circuit and control algorithms have also been evaluated to maximize the power output and to deliver the premium ride comfort and handling performance. Different types of parameterized road excitations have been applied to vehicle suspension systems to investigate the performance of the regenerative shock absorbers. The potential of incorporating nonlinearity into the regenerative shock absorber design analysis is discussed. The research gaps for the comparison of the different drive modes and the nonlinearity analysis of the regenerative shock absorbers are identified and, the corresponding research questions have been proposed for future work.

  15. Inspection of an artificial heart by the neutron radiography technique

    International Nuclear Information System (INIS)

    Pugliesi, R.; Geraldo, L.P.; Andrade, M.L.G.; Menezes, M.O.; Pereira, M.A.S.; Maizato, M.J.S.

    1999-01-01

    The neutron radiography technique was employed to inspect an artificial heart prototype which is being developed to provide blood circulation for patients expecting heart transplant surgery. The radiographs have been obtained by the direct method with a gadolinium converter screen along with the double coated Kodak-AA emulsion film. The artificial heart consists of a flexible plastic membrane located inside a welded metallic cavity, which is employed for blood pumping purposes. The main objective of the present inspection was to identify possible damages in this plastic membrane, produced during the welding process of the metallic cavity. The obtained radiographs were digitized as well as analysed in a PC and the improved images clearly identify several damages in the plastic membrane, suggesting changes in the welding process

  16. Inspection of an artificial heart by the neutron radiography technique

    CERN Document Server

    Pugliesi, R; Andrade, M L G; Menezes, M O; Pereira, M A S; Maizato, M J S

    1999-01-01

    The neutron radiography technique was employed to inspect an artificial heart prototype which is being developed to provide blood circulation for patients expecting heart transplant surgery. The radiographs have been obtained by the direct method with a gadolinium converter screen along with the double coated Kodak-AA emulsion film. The artificial heart consists of a flexible plastic membrane located inside a welded metallic cavity, which is employed for blood pumping purposes. The main objective of the present inspection was to identify possible damages in this plastic membrane, produced during the welding process of the metallic cavity. The obtained radiographs were digitized as well as analysed in a PC and the improved images clearly identify several damages in the plastic membrane, suggesting changes in the welding process.

  17. Morphologic characterization of Mycobacterium marinum by neutron radiographic technique

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jaqueline Michele da; Crispim, Verginia Reis, E-mail: vrcrispim@gmail.com [Universidade Federal do Rio de Janeiro (CT/UFRJ) Centro Tecnologico, Engenharia Nuclear, RJ (Brazil); Silva, Marlei Gomes da [Universidade Federal do Rio de Janeiro (CCS/UFRJ), Centro de Ciencias da Saude. Instituto de Microbiologia Professor Paulo de Goes (Brazil)

    2011-07-01

    The genus Mycobacterium shares many characteristics with the genera Corynebacterium and Actinomyces, among which, similar genome content of bases Guanine-Cytosine and the production of branched long-chain fatty acids called mycolic acids should be enhanced. Mycobacteria are strict aerobic, considered weakly Gram-positive, rod-shaped microorganisms, not possessing flagella. They are intracellular infecting and proliferating in the interior of macrophages, they do not form spores, produce toxins or have capsule. Optimal growth temperature and rate are variable. The genus encompasses approximately 120 known species; however, the present study focuses the characterization of Mycobacterium marinum. This species is generally pathogenic causing deep skin infections. Colonies grow slowly at temperatures around 37 degree C. The aim of this study is to speed the process of M. Marinum morphologic characterization and, in the future, apply it to other species of Nontuberculous Mycobacteria (NTM. In relation to conventional microbiologic essays that usually demand 28 days for colony growth, nuclear testing, using the neutron radiography technique, prove to be much faster. The samples were initially sterilized at the Mycobacteria Laboratory/IMPPG/UFRJ using hypochlorite solution, gluta + formaldehyde and warmed distilled water, according conventional protocols. Then, they were incubated with sodium borate, deposited over CR-39 sheets, fixed with casein (only the first and third sample) and irradiated with a thermal neutron beam generated at the J-9 channel of the Argonauta reactor from the IEN/CNEN. To this end, the following parameters were optimized: incubation time, irradiation time and CR-39 developing time. The images registered in CR-39 were visualized with the help of a Nikon E-400 optical microscope and captured with a Cool pix 995 digital camera. The results showed that the technique produces enlarged images, making it easier the morphologic characterization of

  18. Morphologic characterization of Mycobacterium marinum by neutron radiographic technique

    International Nuclear Information System (INIS)

    Silva, Jaqueline Michele da; Crispim, Verginia Reis; Silva, Marlei Gomes da

    2011-01-01

    The genus Mycobacterium shares many characteristics with the genera Corynebacterium and Actinomyces, among which, similar genome content of bases Guanine-Cytosine and the production of branched long-chain fatty acids called mycolic acids should be enhanced. Mycobacteria are strict aerobic, considered weakly Gram-positive, rod-shaped microorganisms, not possessing flagella. They are intracellular infecting and proliferating in the interior of macrophages, they do not form spores, produce toxins or have capsule. Optimal growth temperature and rate are variable. The genus encompasses approximately 120 known species; however, the present study focuses the characterization of Mycobacterium marinum. This species is generally pathogenic causing deep skin infections. Colonies grow slowly at temperatures around 37 degree C. The aim of this study is to speed the process of M. Marinum morphologic characterization and, in the future, apply it to other species of Nontuberculous Mycobacteria (NTM. In relation to conventional microbiologic essays that usually demand 28 days for colony growth, nuclear testing, using the neutron radiography technique, prove to be much faster. The samples were initially sterilized at the Mycobacteria Laboratory/IMPPG/UFRJ using hypochlorite solution, gluta + formaldehyde and warmed distilled water, according conventional protocols. Then, they were incubated with sodium borate, deposited over CR-39 sheets, fixed with casein (only the first and third sample) and irradiated with a thermal neutron beam generated at the J-9 channel of the Argonauta reactor from the IEN/CNEN. To this end, the following parameters were optimized: incubation time, irradiation time and CR-39 developing time. The images registered in CR-39 were visualized with the help of a Nikon E-400 optical microscope and captured with a Cool pix 995 digital camera. The results showed that the technique produces enlarged images, making it easier the morphologic characterization of

  19. 3-D Deep Penetration Neutron Imaging of Thick Absorbing and Diffusive Objects Using Transport Theory. Final technical report

    International Nuclear Information System (INIS)

    Ragusa, Jean; Bangerth, Wolfgang

    2011-01-01

    here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the

  20. Conceptual design of a two-phase flow absorber system for neutron flux regulation in a CANDU-PHW-1250 reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Moeck, E.O.

    1979-07-01

    A two-phase absorber control (TOPAC) system has been under development at the Chalk River Nuclear Laboratories to meet the need for improved spatial neutron flux control for future CANDU power reactors. Aspects of the conceptual design study presented in this paper include system controllability, in-reactor noise sensitiity, the effect of equipment malfunctions on plant operation, and a comparison with competing systems. The TOPAC system is shown to be a viable alternative to existing and future neutron flux regulating systems based on liquid H 2 O zone compartments. (auth)

  1. Characterization of the neutron sources storage pool of the Neutron Standards Laboratory, using Montecarlo Techniques

    International Nuclear Information System (INIS)

    Campo Blanco, X.

    2015-01-01

    The development of irradiation damage resistant materials is one of the most important open fields in the design of experimental facilities and conceptual nucleoelectric fusion plants. The Neutron Standards Laboratory aims to contribute to this development by allowing the neutron irradiation of materials in its calibration neutron sources storage pool. For this purposes, it is essential to characterize the pool itself in terms of neutron fluence and spectra due to the calibration neutron sources. In this work, the main features of this facility are presented and the characterization of the storage pool is carried out. Finally, an application is shown of the obtained results to the neutron irradiation of material.

  2. {sup 10}B areal density: A novel approach for design and fabrication of B{sub 4}C/6061Al neutron absorbing materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuli [School of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China); Wang, Wenxian, E-mail: wangwenxian@tyut.edu.cn [School of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China); Zhou, Jun [School of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan 030024 (China); Department of Mechanical Engineering, Pennsylvania State University Erie, The Behrend College, Erie, PA 16563 (United States); Chen, Hongsheng [School of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China); Zhang, Peng [School of Materials Science and Engineering, Taiyuan University of Technology, Taiyuan 030024 (China); College of Physics and Optoelectronics, Taiyuan University of Technology, Taiyuan 030024 (China)

    2017-04-15

    In this paper, a novel approach to evaluate the neutron shielding performance of a boron-containing neutron absorbing material was proposed for the first time through the establishment of a direct relationship between {sup 10}B areal density ({sup 10}BAD) of the material and its neutron absorption ratio. It is found when the {sup 10}BAD of a material is greater than 0.034 g/cm{sup 2}, the material will achieve a good neutron shielding performance. Based on this proposed approach, B{sub 4}C/6061Al composite plates with different B{sub 4}C content (10 wt%, 20 wt%, 30 wt%) were successfully fabricated using vacuum hot pressing followed by hot-extrusion. The characteristics of the B{sub 4}C/Al interface were studied in details using transmission electron microscopy (TEM), and the effects of B{sub 4}C particle content on microstructure and mechanical properties of the Al matrix were investigated. Through current studies, B{sub 4}C/6061Al composite plates possessing good neutron shielding performance and tensile strength are found to be able to be fabricated using either 20 wt% of B{sub 4}C content with a plate thickness of 4.5 mm or 30 wt% B{sub 4}C content with a plate thickness of 3 mm. - Graphical abstract: In this paper, a novel approach to evaluate the neutron shielding ability of a boron-containing neutron shielding material was proposed for the first time through the establishment of a direct relationship between {sup 10}B area density ({sup 10}BAD) of the material and its neutron shielding ratio. - Highlights: •{sup 10}BAD was proposed to evaluate the boron-containing neutron absorber material’s neutron shielding performance. •The direct relationship between the {sup 10}BAD and neutron shielding performance was firstly established. •TEM analysis of the composites reveals that an amorphous layer exists at the Al/B{sub 4}C interface. •Suitable B{sub 4}C contents and thickness for the fabrication of B{sub 4}C/6061A1 NAC plate were given in the

  3. 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

  4. Hot deformation behaviors and processing maps of B{sub 4}C/Al6061 neutron absorber composites

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yu-Li [School of Materials Science and Engineering, Taiyuan University Of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China); Wang, Wen-Xian, E-mail: Wangwenxian@tyut.edu.cn [School of Materials Science and Engineering, Taiyuan University Of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China); Zhou, Jun [School of Materials Science and Engineering, Taiyuan University Of Technology, Taiyuan 030024 (China); Department of Mechanical Engineering, Pennsylvania State University Erie, The Behrend College, Erie, PA 16563 (United States); Chen, Hong-Sheng [School of Materials Science and Engineering, Taiyuan University Of Technology, Taiyuan 030024 (China); Key Laboratory of Interface Science and Engineering in Advanced Materials, Ministry of Education, Taiyuan University of Technology, Taiyuan 030024 (China)

    2017-02-15

    In this study, the hot deformation behaviors of 30 wt.% B{sub 4}C/Al6061 neutron absorber composites (NACs) have been investigated by conducting isothermal compression tests at temperatures ranging from 653 K to 803 K and strain rates from 0.01 to 10 s{sup −1}. It was found that, during hot compression, the B{sub 4}C/Al6061 NACs exhibited a steady flow characteristic which can be expressed by the Zener-Hollomon parameter as a hyperbolic-sine function of flow stress. High average activation energy (185.62 kJ/mol) of B{sub 4}C/Al6061 NACs is noted in current study owing to the high content of B{sub 4}C particle. The optimum hot working conditions for B{sub 4}C/Al6061 NACs are found to be 760–803 K/0.01–0.05 s{sup −1} based on processing map and microstructure evolution. Typical material instabilities are thought to be attributed to void formation, adiabatic shear bands (ASB), particle debonding, and matrix cracking. Finally, the effect of the plastic deformation zones (PDZs) on the microstructure evolution in this 30 wt.% B{sub 4}C/Al6061 composite is found to be very important. - Highlights: •The hot deformation behavior of the 30 wt.% B{sub 4}C/Al6061 NACs was first analyzed. •The 3D efficiency map and the instability map are developed. •The optimum hot working conditions were identified and validated by SEM and TEM. •The hot deformation schematic diagram of 30 wt.% B{sub 4}C/Al6061 NACs is developed.

  5. Characterization and MCNP simulation of neutron energy spectrum shift after transmission through strong absorbing materials and its impact on tomography reconstructed image.

    Science.gov (United States)

    Hachouf, N; Kharfi, F; Boucenna, A

    2012-10-01

    An ideal neutron radiograph, for quantification and 3D tomographic image reconstruction, should be a transmission image which exactly obeys to the exponential attenuation law of a monochromatic neutron beam. There are many reasons for which this assumption does not hold for high neutron absorbing materials. The main deviations from the ideal are due essentially to neutron beam hardening effect. The main challenges of this work are the characterization of neutron transmission through boron enriched steel materials and the observation of beam hardening. Then, in our work, the influence of beam hardening effect on neutron tomographic image, for samples based on these materials, is studied. MCNP and FBP simulation are performed to adjust linear attenuation coefficients data and to perform 2D tomographic image reconstruction with and without beam hardening corrections. A beam hardening correction procedure is developed and applied based on qualitative and quantitative analyses of the projections data. Results from original and corrected 2D reconstructed images obtained shows the efficiency of the proposed correction procedure. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Small-angle neutron scattering technique in liquid crystal studies

    International Nuclear Information System (INIS)

    Shahidan Radiman

    2005-01-01

    The following topics discussed: general principles of SAS (Small-angle Neutron Scattering), liquid crystals, nanoparticle templating on liquid crystals, examples of SAS results, prospects of this studies

  7. Neutron diagnostics on TFTR utilizing the Campbelling technique

    International Nuclear Information System (INIS)

    England, A.C.; Hendel, H.W.; Neischmidt, E.B.

    1986-01-01

    The authors report modified commercial neutron counting equipment installed on a tokamak fusion test reactor (TFTR) which utilizes the Campbelling theorem to monitor the neutron source strength at very high neutron count rates. Campbelling utilizes the large amplitude fluctuation from neutron events in the detectors to discriminate against small amplitude noise events. Source strengths yielding equivalent count rates a factor of five higher than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated

  8. Radiolysis of some aqueous solutions of neutron absorbers; Etude des effets de certains absorbeurs de neutrons en solution sur la radiolyse de l'eau

    Energy Technology Data Exchange (ETDEWEB)

    Rozenberg, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-12-15

    The initial yield of molecular hydrogen formed by radiolytic decomposition of water in reactor and {sup 60}Co gamma radiation is decreased by the presence of salts of polyvalent elements possessing only one stable valence, i.e cadmium, zinc, magnesium, gadolinium. This effect is favourable for the use of cadmium and gadolinium as soluble neutron absorber in heavy water reactors. Cations of these salts are not inert toward the primary products of water radiolysis. They have a high degree of reactivity toward the hydrated electron, which is the precursor of molecular hydrogen in neutral or alkaline aqueous media. The value of the rate constant for the reaction between cadmium ion and hydrated electron was shown to be (6.1 {+-} 1.8) 10{sup 10} M{sup -1} s{sup -1}. Boric acid at low concentration has no effect on the radiation chemistry of water. An isotope effect has been found in the radiolysis of heavy water, corresponding to a lowering of initial yield [G{sub 0}(D{sub 2}) < G{sub 0}(H{sub 2})]. additionally it was necessary to determine the influence of organic impurities, remaining after the purification of water, on the mechanism of its radiolytic decomposition. (author) [French] Le rendement initial de la formation d'hydrogene moleculaire dans la decomposition radiolytique de l'eau, sous l'effet du rayonnement des reacteurs nucleaires ou du cobalt 60, est diminue si le solute est un sel d'element polyvalent ne possedant qu'un seul etat stable de valence (cadmium, zinc, magnesium, gadolinium). Cet effet est favorable au choix des elements cadmium et gadolinium pour servir d'absorbeur soluble de neutrons dans un reacteur a eau lourde. Les cations de ces sels ne sont pas inertes vis-a-vis des produits primaires de la radiolyse. Ils ont une affinite notable pour l'electron solvate, precurseur de l'hydrogene moleculaire en milieu neutre ou alcalin. En particulier, la constante de vitesse de la reaction du cadmium ionise avec l'electron solvate a pu etre calculee. Sa

  9. High-sensitivity measurements for low-level TRU wastes using advanced passive neutron techniques

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.W.

    1992-01-01

    In recent years, both passive- and active-neutron nondestructive assay (NDA) systems have been used to measure the uranium and plutonium content in 200-ell drums. Because of the heterogeneity of the wastes, representative sampling is not possible and NDA methods are preferred over destructive analysis. Active-neutron assay systems are used to measure the fissile isotopes such as 235 U, 23 Pu, and 241 Pu; the isotopic ratios are used to infer the total plutonium content and thus the specific disintegration rate. The active systems include 14-MeV-neutron (DT) generators with delayed-neutron counting, (D,T) generators with the differential die-away technique, and 252 Cf delayed-neutron shufflers. Passive assay systems (for example, segmented gamma-ray scanners)5 have used gamma-ray sessions, while others (for example, passive drum counters) used passive-neutron signals. We have developed a new passive-neutron measurement technique to improve the accuracy and sensitivity of the NDA of plutonium scrap and waste. This new 200-ell-drum assay system combines the classical NDA method of counting passive-neutron totals and coincidences from plutonium with the new features of ''add-a-source'' (AS) and multiplicity counting to improve the accuracy of matrix corrections and statistical techniques that improve the low-level detectability limits. This paper describes the improvements we have made in passive-neutron assay systems and compares the accuracies and detectability limits of passive- and active-neutron assay systems

  10. Neutron absorber pellets

    International Nuclear Information System (INIS)

    Radford, K.C.

    1983-01-01

    An annular burnable poison pellet of aluminium oxide - boron carbide (Al 2 O 3 - B 4 C) adapted for positioning in the annular space of concentrically disposed zircaloy tubes. Each tubular pellet is fabricated from Al 2 O 3 powders of moderate sintering activity which serves as a matrix for B 4 C medium size particle distribution. Special pellet moisture controls are incorporated in the pellet for moisture stability and the pellet is sintered in the temperature range of 1630 deg to 1650 deg C. This method of fabrication produces a pellet about 2 inch long with a wall thickness of from 0.020 inch to 0.040 inch. Fabricating each pellet to about 70% theoretical density gives an optimum compromise between fabricability, microstructure, strength and moisture absorption. (author)

  11. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1985-01-01

    This invention provides ceramic processing including sintering schedules which produce annular pellets containing burnable poisons for use in reactor control rods. Typically the powder includes Al 2 O 3 and from 1 to 50 weight percent B 4 C. The Al 2 O 3 and B 4 C, appropriately sized, are milled in a ball mill with liquid to produce a slurry. The slurry is spray dried to produce small spheres of the mixed powder, which is mixed with adequate organic binder and plasticizer and formed into a hollow green body having the shape of a tube. The green body is sintered to produce a ceramic tube from which the pellets are cut. The tube is sintered to size so that the pellets have the required dimensions. It is an important feature of this invention that the powder is formed into the green body by applying isostatic pressure to the powder

  12. Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

  13. Designing on-line analyzer for coal on belt conveyor using neutron activation technique

    International Nuclear Information System (INIS)

    Rony Djokorayono; Agus Cahyono

    2014-01-01

    Basic design of on-line analyzer for coal on belt conveyor using neutron activation technique has been carried out. Compared with sampling technique, this neutron activation technique has some advantages in term of analysis accuracy and time. The design activities performed include the establishment of design requirements, functional requirements, technical requirements, technical specification, detection sub-system design, data acquisition subsystem design, and operator computer console design. This program will use Nal(Tl) scintillation detector to detect gamma-rays emitted by elements in coal due to neutron activation of a neutron source, "2"5"2Cf (Californium-252). This basic design of on-line analyzer for coal on belt conveyor using neutron activation technique should be followed up with the development of detailed design, prototype construction, and field testing. (author)

  14. Investigation of Non-Vacuum Deposition Techniques in Fabrication of Chalcogenide-Based Solar Cell Absorbers

    KAUST Repository

    Alsaggaf, Ahmed

    2015-01-01

    simulating R2R process design at speeds up to 1.05 m/min. The research clarified morphological and compositional impacts of Nd:YAG laser heat-treatment on Cu(In,Ga)Se2 absorber layer to simplify the annealing step in non-vacuum environment compatible to R2R

  15. A technique for determining the deuterium/hydrogen contrast map in neutron macromolecular crystallography.

    Science.gov (United States)

    Chatake, Toshiyuki; Fujiwara, Satoru

    2016-01-01

    A difference in the neutron scattering length between hydrogen and deuterium leads to a high density contrast in neutron Fourier maps. In this study, a technique for determining the deuterium/hydrogen (D/H) contrast map in neutron macromolecular crystallography is developed and evaluated using ribonuclease A. The contrast map between the D2O-solvent and H2O-solvent crystals is calculated in real space, rather than in reciprocal space as performed in previous neutron D/H contrast crystallography. The present technique can thus utilize all of the amplitudes of the neutron structure factors for both D2O-solvent and H2O-solvent crystals. The neutron D/H contrast maps clearly demonstrate the powerful detectability of H/D exchange in proteins. In fact, alternative protonation states and alternative conformations of hydroxyl groups are observed at medium resolution (1.8 Å). Moreover, water molecules can be categorized into three types according to their tendency towards rotational disorder. These results directly indicate improvement in the neutron crystal structure analysis. This technique is suitable for incorporation into the standard structure-determination process used in neutron protein crystallography; consequently, more precise and efficient determination of the D-atom positions is possible using a combination of this D/H contrast technique and standard neutron structure-determination protocols.

  16. TLD-300 detectors for separate measurement of total and gamma absorbed dose distributions of single, multiple, and moving-field neutron treatments

    International Nuclear Information System (INIS)

    Rassow, J.

    1984-01-01

    Fast neutron therapy requirements, because of the poor depth dose characteristic of present therapeutical sources, are at least as complex in treatment plans as photon therapy. The physical part of the treatment planning is very important; however, it is much more complicated than for photons or electrons owing to the need for: Separation of total and gamma absorbed dose distributions (Dsub(T) and Dsub(G)); and more stringent tissue-equivalence conditions of phantoms than in photon therapy. Therefore, methods of clinical dosimetry for the separate determination of total and gamma absorbed dose distributions in irregularly shaped (inhomogeneous) phantoms are needed. A method using TLD-300 (CaF 2 :Tm) detectors is described, which is able to give an approximate solution of the above-mentioned dosimetric requirements. The two independent doses, Dsub(T) and Dsub(G), can be calculated by an on-line computer analysis of the digitalized glow curve of TLD-300 detectors, irradiated with d(14)+Be neutrons of the cyclotron isocentric neutron therapy facility CIRCE in Essen. Results are presented for depth and lateral absorbed dose distributions (Dsub(T) and Dsub(G)) for fixed neutron beams of different field sizes compared with measurements by standard procedures (TE-TE ionization chamber, GM counter) in an A-150 phantom. The TLD-300 results for multiple and moving-field treatments (with and without wedge filters) in a patient simulating irregularly shaped (inhomogeneous) phantoms, are shown together with computer calculations of these dose distributions. The probable causes for some systematic deviations are discussed, which lead to open problems for further investigations owing to features of the detector material and the evaluation method, but mainly to differences in the composition of phantom materials used for the calculations (standard dose distributions) and TLD-300 measurements. (author)

  17. Monte Carlo Simulation of the Time-Of-Flight Technique for the Measurement of Neutron Cross-section in the Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.

  18. Experimental technique of stress analyses by neutron diffraction

    International Nuclear Information System (INIS)

    Sun, Guangai; Chen, Bo; Huang, Chaoqiang

    2009-09-01

    The structures and main components of neutron diffraction stress analyses spectrometer, SALSA, as well as functions and parameters of each components are presented. The technical characteristic and structure parameters of SALSA are described. Based on these aspects, the choice of gauge volume, method of positioning sample, determination of diffraction plane and measurement of zero stress do are discussed. Combined with the practical experiments, the basic experimental measurement and the related settings are introduced, including the adjustments of components, pattern scattering, data recording and checking etc. The above can be an instruction for stress analyses experiments by neutron diffraction and neutron stress spectrometer construction. (authors)

  19. Improved cable compensation technique for self powered neutron detectors

    International Nuclear Information System (INIS)

    Nieuwenhove, R. van

    1996-01-01

    Measurements with cobalt self powered neutron detectors on the BR2 reactor have revealed that the currents induced by external gamma radiation can be of the same order as the neutron induced signal and that the gamma induced current on the emitter and the compensator wires are not symmetric. In this case, the standard detection electronic setup leads to erroneous results. It is shown that a slightly modified electronic setup, in which this asymmetry is compensated for, can nevertheless allow to obtain correct neutron flux measurements. Measures to reduce the influence of external gamma radiation in general will also be discussed. (orig.)

  20. Neutron Spectroscopy for pulsed beams with frame overlap using a double time-of-flight technique

    Science.gov (United States)

    Harrig, K. P.; Goldblum, B. L.; Brown, J. A.; Bleuel, D. L.; Bernstein, L. A.; Bevins, J.; Harasty, M.; Laplace, T. A.; Matthews, E. F.

    2018-01-01

    A new double time-of-flight (dTOF) neutron spectroscopy technique has been developed for pulsed broad spectrum sources with a duty cycle that results in frame overlap, where fast neutrons from a given pulse overtake slower neutrons from previous pulses. Using a tunable beam at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory, neutrons were produced via thick-target breakup of 16 MeV deuterons on a beryllium target in the cyclotron vault. The breakup spectral shape was deduced from a dTOF measurement using an array of EJ-309 organic liquid scintillators. Simulation of the neutron detection efficiency of the scintillator array was performed using both GEANT4 and MCNP6. The efficiency-corrected spectral shape was normalized using a foil activation technique to obtain the energy-dependent flux of the neutron beam at zero degrees with respect to the incoming deuteron beam. The dTOF neutron spectrum was compared to spectra obtained using HEPROW and GRAVEL pulse height spectrum unfolding techniques. While the unfolding and dTOF results exhibit some discrepancies in shape, the integrated flux values agree within two standard deviations. This method obviates neutron time-of-flight spectroscopy challenges posed by pulsed beams with frame overlap and opens new opportunities for pulsed white neutron source facilities.

  1. Absorbed dose to water determination with ionization chamber dosimetry and calorimetry in restricted neutron, photon, proton and heavy-ion radiation fields.

    Science.gov (United States)

    Brede, H J; Greif, K-D; Hecker, O; Heeg, P; Heese, J; Jones, D T L; Kluge, H; Schardt, D

    2006-08-07

    Absolute dose measurements with a transportable water calorimeter and ionization chambers were performed at a water depth of 20 mm in four different types of radiation fields, for a collimated (60)Co photon beam, for a collimated neutron beam with a fluence-averaged mean energy of 5.25 MeV, for collimated proton beams with mean energies of 36 MeV and 182 MeV at the measuring position, and for a (12)C ion beam in a scanned mode with an energy per atomic mass of 430 MeV u(-1). The ionization chambers actually used were calibrated in units of air kerma in the photon reference field of the PTB and in units of absorbed dose to water for a Farmer-type chamber at GSI. The absorbed dose to water inferred from calorimetry was compared with the dose derived from ionometry by applying the radiation-field-dependent parameters. For neutrons, the quantities of the ICRU Report 45, for protons the quantities of the ICRU Report 59 and for the (12)C ion beam, the recommended values of the International Atomic Energy Agency (IAEA) protocol (TRS 398) were applied. The mean values of the absolute absorbed dose to water obtained with these two independent methods agreed within the standard uncertainty (k = 1) of 1.8% for calorimetry and of 3.0% for ionometry for all types and energies of the radiation beams used in this comparison.

  2. Nuclear analytical techniques with neutron beams at the Univ. of Texas at Austin

    International Nuclear Information System (INIS)

    Uenlue, K.; Wehring, B.W.

    1996-01-01

    Neutron beams produced by nuclear research reactors can be used for analytical chemical analysis by measuring nuclear radiation produced by neutron capture. Prompt gamma activation analysis (PGAA) and neutron depth profiling (NDP) are two such analytical techniques. For the last three decades, these techniques have been applied at a number of research reactors around the world. Within the last 4 yr, we have developed NDP and PGAA facilities at The University of Texas at Austin research reactor, a 1-MW TRIGA Mark II reactor. Brief descriptions of the facilities and summaries of activities for these analytical techniques at the University of Texas at Austin are provided in this paper

  3. The new generations of power components will depend on neutron and/or electron bombardment techniques

    International Nuclear Information System (INIS)

    Lilen, H.

    1976-01-01

    Neutron and electron bombardment techniques for materials doping, newly introduced in the fabrication of power semiconductor components: diodes, transistors, thyristors, and triacs are briefly outlined. A neutron bombardment of high purity silicon results in a short-lived 31 Si isotope (from 30 Si) decaying into 31 P. The phosphorus with its five peripheral electrons induces a negative doping (N), and the neutron technique gives a homogeneous doping. Furthermore, silicon bombardment with 1 to 2MeV electrons induces micro-ruptures in the lattice, that act as recombination traps reducing carrier lifetimes. Consequently, gold diffusion techniques can be replaced by electron bombardment with a gain in controlling carrier lifetimes [fr

  4. Advancement of neutron radiography technique in JRR-3M

    International Nuclear Information System (INIS)

    Matsubayashi, Masahito

    1999-01-01

    The JRR-3M thermal neutron radiography facility (JRR-3M TNRF) was completed in the JRR-3M of the Japan Atomic Energy Research Institute in 1991 and has been utilized as research tools for various kinds of research fields such as thermal hydraulic researches, agricultural researches, medical researches, archaeological researches and so on. High performance of the JRR-3M TNRF such as high neutron flux, high collimator ratio and wide radiographing field has enabled advanced researches and stimulated developments of advanced neutron radiography (NR) systems for higher spatial resolution and for higher temporal resolution. Static NR systems using neutron imaging plates or cooled CCD camera with high spatial resolution, a real-time NR system using a silicon intensifier target tube camera and a high-frame-rate NR system using a combination of an image intensifier and a high speed digital video camera with high temporal resolution have been developed to fill the requirements from researchers. (author)

  5. Neutron diagnostics on TFTR utilizing the Campbelling technique

    International Nuclear Information System (INIS)

    England, A.C.; Hendel, H.W.; Nieschmidt, E.B.

    1986-01-01

    Modified commercial equipment installed on the tokamak fusion test reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) utilizes Campbell's mean square voltage theorem to monitor the neutron source strength at neutron count rates orders of magnitude above the capability of the count rate mode. Campbelling uses the large amplitude fluctuations from neutron fission events in the detectors to discriminate against small amplitude γ ray and other noise events. Source strengths yielding equivalent count rates a factor of 5 greater than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated. Fundamental advantages are the extended useful range of the detectors by a factor of --10 4 and gamma rejection by a factor of --10 3 . Some results are shown and the neutron source strengths obtained are compared to those from conventional counting circuits and from other detectors whose outputs have not yet suffered counting losses

  6. Improved mesh based photon sampling techniques for neutron activation analysis

    International Nuclear Information System (INIS)

    Relson, E.; Wilson, P. P. H.; Biondo, E. D.

    2013-01-01

    The design of fusion power systems requires analysis of neutron activation of large, complex volumes, and the resulting particles emitted from these volumes. Structured mesh-based discretization of these problems allows for improved modeling in these activation analysis problems. Finer discretization of these problems results in large computational costs, which drives the investigation of more efficient methods. Within an ad hoc subroutine of the Monte Carlo transport code MCNP, we implement sampling of voxels and photon energies for volumetric sources using the alias method. The alias method enables efficient sampling of a discrete probability distribution, and operates in 0(1) time, whereas the simpler direct discrete method requires 0(log(n)) time. By using the alias method, voxel sampling becomes a viable alternative to sampling space with the 0(1) approach of uniformly sampling the problem volume. Additionally, with voxel sampling it is straightforward to introduce biasing of volumetric sources, and we implement this biasing of voxels as an additional variance reduction technique that can be applied. We verify our implementation and compare the alias method, with and without biasing, to direct discrete sampling of voxels, and to uniform sampling. We study the behavior of source biasing in a second set of tests and find trends between improvements and source shape, material, and material density. Overall, however, the magnitude of improvements from source biasing appears to be limited. Future work will benefit from the implementation of efficient voxel sampling - particularly with conformal unstructured meshes where the uniform sampling approach cannot be applied. (authors)

  7. Neutron-based techniques for detection of explosives and drugs

    Energy Technology Data Exchange (ETDEWEB)

    Kiraly, B.; Olah, L.; Csikai, J. E-mail: csikai@falcon.phys.klte.hu

    2001-06-01

    Systematic measurements were carried out on the possible use of elastically backscattered Pu-Be neutrons combined with the thermal neutron reflection method for the identification of land mines and illicit drugs via he detection of H, C, N, and O elements as their major constituents. While ur present results show that these methods are capable of indicating the anomalies in bulky materials and observation of the major elements, e termination of the exact atom fractions needs further investigation.

  8. Neutron-based techniques for detection of explosives and drugs

    CERN Document Server

    Kiraly, B; Csikai, J

    2001-01-01

    Systematic measurements were carried out on the possible use of elastically backscattered Pu-Be neutrons combined with the thermal neutron reflection method for the identification of land mines and illicit drugs via he detection of H, C, N, and O elements as their major constituents. While ur present results show that these methods are capable of indicating the anomalies in bulky materials and observation of the major elements, e termination of the exact atom fractions needs further investigation.

  9. The neutron spin-echo spectrometer: a new high resolution technique in neutron scattering

    International Nuclear Information System (INIS)

    Nicholson, L.K.

    1981-01-01

    The neutron spin-echo (NSE) spectrometer provides the highest energy resolution available in neutron scattering experiments. The article describes the principles behind the first NSE spectrometer (at the Institute Laue-Langevin, Grenoble, France) and, as an example of one of its applications, some recent results on polymer chain dynamics are presented. (author)

  10. New experimental research stand SVICKA neutron field analysis using neutron activation detector technique

    Science.gov (United States)

    Varmuza, Jan; Katovsky, Karel; Zeman, Miroslav; Stastny, Ondrej; Haysak, Ivan; Holomb, Robert

    2018-04-01

    Knowledge of neutron energy spectra is very important because neutrons with various energies have a different material impact or a biological tissue impact. This paper presents basic results of the neutron flux distribution inside the new experimental research stand SVICKA which is located at Brno University of Technology in Brno, Czech Republic. The experiment also focused on the investigation of the sandwich biological shielding quality that protects staff against radiation effects. The set of indium activation detectors was used to the investigation of neutron flux distribution. The results of the measurement provide basic information about the neutron flux distribution inside all irradiation channels and no damage or cracks are present in the experimental research stand biological shielding.

  11. Electron-volt spectroscopy at a pulsed neutron source using a resonance detector technique

    CERN Document Server

    Andreani, C; Senesi, R; Gorini, G; Tardocchi, M; Bracco, A; Rhodes, N; Schooneveld, E M

    2002-01-01

    The effectiveness of the neutron resonance detector spectrometer for deep inelastic neutron scattering measurements has been assessed by measuring the Pb scattering on the eVS spectrometer at ISIS pulsed neutron source and natural U foils as (n,gamma) resonance converters. A conventional NaI scintillator with massive shielding has been used as gamma detector. A neutron energy window up to 90 eV, including four distinct resonance peaks, has been assessed. A net decrease of the intrinsic width of the 6.6 eV resonance peak has also been demonstrated employing the double difference spectrum technique, with two uranium foils of different thickness.

  12. A high-resolution neutron spectra unfolding method using the Genetic Algorithm technique

    CERN Document Server

    Mukherjee, B

    2002-01-01

    The Bonner sphere spectrometers (BSS) are commonly used to determine the neutron spectra within various nuclear facilities. Sophisticated mathematical tools are used to unfold the neutron energy distribution from the output data of the BSS. This paper highlights a novel high-resolution neutron spectra-unfolding method using the Genetic Algorithm (GA) technique. The GA imitates the biological evolution process prevailing in the nature to solve complex optimisation problems. The GA method was utilised to evaluate the neutron energy distribution, average energy, fluence and equivalent dose rates at important work places of a DIDO class research reactor and a high-energy superconducting heavy ion cyclotron. The spectrometer was calibrated with a sup 2 sup 4 sup 1 Am/Be (alpha,n) neutron standard source. The results of the GA method agreed satisfactorily with the results obtained by using the well-known BUNKI neutron spectra unfolding code.

  13. 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)

  14. Overlapping bio-absorbable scaffolds: Aim for D2D technique?

    Science.gov (United States)

    Khan, Asaad A; Dangas, George D

    2018-06-01

    The results of overlapping metallic stents have been concerning but this practice is often unavoidable in the setting of long or tortuous lesions, diameter discrepancy of proximal and distal vessel, and for residual dissections. Theoretically, bio-absorbable scaffolds may carry an advantage over metallic stents due to the progressive resorption of the scaffold theoretically rendering the overlap a non-issue; this has not been clinically evident. Since stent/scaffold overlap cannot be entirely avoided, improved stent delivery/deployment and scaffold design modification may reduce complications in this complex patient subset. © 2018 Wiley Periodicals, Inc.

  15. 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

  16. Determination Of Natural Boron Concentration In Coffee Leaves, Using de Autobiography by Neutron Capture Technique

    International Nuclear Information System (INIS)

    Loria, L. G.; Jimenez, R.; Thellier, M.

    1999-01-01

    Determination of natural boron concentration in coffee leaves, using the autoradiography, by neutron capture technique. The boron absorption coefficient in young coffee leaves was measured using autoradiography by neutron capture. In two experiments carried out in April and November, 1996, it was found that the coefficient varies between 0.9 and 5.3 nmol/h. the concentration of natural boron in coffee leaves in regard to age, symptoms and treatment received was also studied, using the same technique. (Author) [es

  17. Neutron-based techniques for detection of explosives and drugs

    International Nuclear Information System (INIS)

    Kiraly, B.; Olah, L.; Csikai, G.J.

    2000-01-01

    Neutron reflection, scattering and transmission methods combined with the detection of characteristic gamma rays have an increasing role in the identification of hidden explosives, illicit drugs and other contraband materials. There are about 100 million land mines buried in some 70 countries. Among the abandoned anti-personnel land mines (APL) certain types have low mass (about 100 g) and contain little or no metal. Therefore, these plastic APL cannot be detected by the usual metal detectors. The IAEA Physics Section has organized a CRP in 1999 for the development of novel methods in order to speed up the removing process of APL. The transportation of illicit drugs has shown an increasing trend during the last decade. Developments of fast, non-destructive interrogation methods are required for the inspection of cargo containers, trucks and airline baggage. The major constituents of plastic APL and drugs are H, C, N and O which can be identified by the different neutron interactions. The atom fractions of these elements, in particular the C/O, C/N and C/H ratios, are quite different for drugs and explosives as compared to other materials used to hide them. Recently, we have carried out systematic measurements and calculations on the neutron fields from the 9 Be(d,n), 2 H(d,n), 252 Cf and Pu-Be sources passing through different bulky samples, on the possible use of elastically backscattered Pu-Be neutrons in elemental analysis and on the advantages and limitations of the thermal neutron reflection method in the identification of land mines and illicit drugs. The measured spectral shapes of neutrons were compared with the calculated results using the MCNP-4A and MCNP-4B codes. (author)

  18. Neutron activation: an invaluable technique for teaching applied radiation

    International Nuclear Information System (INIS)

    Trainer, Matthew

    2002-01-01

    This experiment introduces students to the important method of neutron activation. A sample of aluminium was irradiated with neutrons from an isotropic 241 Am-Be source. Using γ-ray spectroscopy, two radionuclide products were identified as 27 Mg and 28 Al. Applying a cadmium cut-off filter and an optimum irradiation time of 45 min, the half-life of 27 Mg was determined as 9.46±0.50 min. The half-life of the 28 Al radionuclide was determined as 2.28±0.10 min using a polythene moderator and an optimum irradiation time of 10 min. (author)

  19. Evaluation of spectral unfolding techniques for neutron spectroscopy

    International Nuclear Information System (INIS)

    Sunden, Erik Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Sjoestrand, H.; Weiszflog, M.; Kaellne, J.; Gorini, G.; Tardocchi, M.

    2008-01-01

    The precision of the JET installations of MAXED, GRAVEL and the L-curve version of MAXED has been evaluated by using synthetic neutron spectra. We have determined the number of counts needed for the detector systems NE213 and MPR to get an error below 10% of the MAXED unfolded neutron spectra is determined to be ∼10 6 and ∼10 4 , respectively. For GRAVEL the same number is ∼10 7 and ∼3·10 4 for NE213 and MPR, respectively

  20. Actinide neutron induced cross section measurements using the oscillation technique in the Minerve reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, B.; Leconte, P.; Gruel, A.; Antony, M.; Di-Salvo, J.; Hudelot, J.P.; Pepino, A.; Lecluze, A. [CEA Cadarache, DEN/CAD/DER/SPRC/LEPh, 13 - Saint-Paul-lez-Durance (France)

    2009-07-01

    CEA is deeply involved research programs concerning nuclear fuel advanced studies (actinides, plutonium), waste management, the scientific and technical support of French PWR reactors and EPR reactor, and innovative systems. In this framework, specific neutron integral experiments have been carried out in the critical ZPR (zero power reactor) facilities of the CEA at Cadarache such as MINERVE, EOLE and MASURCA. This paper deals with MINERVE Pool Reactor experiments. MINERVE is mainly devoted to neutronics studies of different reactor core types. The aim is to improve the knowledge of the integral absorption cross sections of actinides (OSMOSE program), of new absorbers (OCEAN program) and also for fission Products (CBU program) in thermal, epithermal and fast neutron spectra. (authors)

  1. Neutron diffraction measurements at the INES diffractometer using a neutron radiative capture based counting technique

    Energy Technology Data Exchange (ETDEWEB)

    Festa, G. [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Pietropaolo, A., E-mail: antonino.pietropaolo@roma2.infn.it [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Grazzi, F.; Barzagli, E. [CNR-ISC Firenze (Italy); Scherillo, A. [CNR-ISC Firenze (Italy); ISIS facility Rutherford Appleton Laboratory (United Kingdom); Schooneveld, E.M. [ISIS facility Rutherford Appleton Laboratory (United Kingdom)

    2011-10-21

    The global shortage of {sup 3}He gas is an issue to be addressed in neutron detection. In the context of the research and development activity related to the replacement of {sup 3}He for neutron counting systems, neutron diffraction measurements performed on the INES beam line at the ISIS pulsed spallation neutron source are presented. For these measurements two different neutron counting devices have been used: a 20 bar pressure squashed {sup 3}He tube and a Yttrium-Aluminum-Perovskite scintillation detector. The scintillation detector was coupled to a cadmium sheet that registers the prompt radiative capture gamma rays generated by the (n,{gamma}) nuclear reactions occurring in cadmium. The assessment of the scintillator based counting system was done by performing a Rietveld refinement analysis on the diffraction pattern from an ancient Japanese blade and comparing the results with those obtained by a {sup 3}He tube placed at the same angular position. The results obtained demonstrate the considerable potential of the proposed counting approach based on the radiative capture gamma rays at spallation neutron sources.

  2. Air pollution studies in Tianjing city using neutron activation analysis techniques

    International Nuclear Information System (INIS)

    Ni Bangfa; Tian Weizhi; Nie Nuiling; Wang Pingsheng

    1999-01-01

    Two sites of airborne sampling from industrial and residential areas were made in Tianjing city during February and June using PM-10 sampler and analyzed by NAA techniques; Comparison of air pollution between urban and rural area in Tianjing city was made using neutron activation analysis techniques and some other data analyzing techniques. (author)

  3. Status report on the development of a prompt fission neutron uranium borehole logging technique

    International Nuclear Information System (INIS)

    Smith, G.W.

    1977-05-01

    The prompt fission neutron (PFN) method of direct uranium measurement was studied. The PFN uranium logging technique measures the enhanced epithermal neutron population created by the prompt thermal fission of 235 U to assay uranium mineralization around a borehole. This neutron population exists for several hundred microseconds after a pulsed neutron source produces a burst of high energy (14 MeV) neutrons. A feasibility study established the basic relationship between the uranium concentration and the enhanced epithermal neutron count, and defined the major measurement perturbing factors. Following the feasibility study, development of a PFN prototype field probe was undertaken. A laboratory type neutron generator, the Controlatron, was modified for use in the probe. Field evaluation of the prototype system began in January 1976. Comparisons of neutron logs and natural gamma logs taken during this evaluation period clearly define many disequilibrium conditions as verified by ore grade estimates from core samples. The feasibility of the PFN logging technique to detect uranium in-situ has now been demonstrated

  4. Nondestructive analysis of the natural uranium mass through the measurement of delayed neutrons using the technique of pulsed neutron source

    International Nuclear Information System (INIS)

    Coelho, Paulo Rogerio Pinto

    1979-01-01

    This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) 4 He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)

  5. Computational Modeling of a Time-Independent, Heterogeneous Reactor Core Using Simplified Discrete Ordinates Neutron Transport Techniques

    National Research Council Canada - National Science Library

    Labowski, Kristofer

    2001-01-01

    The Linear Characteristic (LC) method on rectangular boxoid meshes is a discrete ordinate neutron transport technique that uses both zeroth and first moments of the angular neutron flux to construct a relatively accurate...

  6. Feasibility study for the investigation of Nitinol self-expanding stents by neutron techniques

    International Nuclear Information System (INIS)

    Rogante, M.; Pasquini, U.; Rosta, L.; Lebedev, V.

    2011-01-01

    In this paper, neutron techniques - in particular, small angle neutron scattering (SANS) and neutron diffraction (ND) - are considered for the non-destructive characterization of Nitinol artery stents. This roughly equiatomic (50Ni-50Ti at%) shape memory alloy (SMA) exhibits significant properties of superelasticity and biocompatibility that make it suitable to be typically used as smart material for medical implants and devices. Nitinol self-expanding artery stents, as permanent vascular support structures, supply an ideal option to bypass surgery, but they are submitted for the whole of patient's life to the dynamical stress of the artery pulsation and the aggression from the biological environment. These stents, consequently, can suffer from wear and fracture occurrence likely due to a variety of cyclic fatigue, overload conditions and residual stresses. Neutrons have recently become a progressively more important probe for various materials and components and they allow achieving information complementary to those obtained from the traditional microstructural analyses. The outputs from the preliminary works already carried out in this field consent to consider neutron techniques capable to contribute to the development of these crucial medical implants. The achievable results can yield trends adoptable in monitoring of the stent features. -- Research Highlights: → Neutron techniques can contribute to develop Nitinol self-expanding artery stents. → Neutrons investigations can help avoiding wear and fracture events in Nitinol stents. → Neutron techniques can yield trends adoptable in monitoring of Nitinol stent features. → SANS is able to perform a micro- and nano-scale characterization of Nitinol stents. → Neutron Diffraction helps assessing stresses due to the exercise in Nitinol stents.

  7. Dosimetric evaluation of the Fricke gel dosimeter using the spectrophotometric technique for application in electron and neutron dosimetry

    International Nuclear Information System (INIS)

    Mangueira, Thyago Fressatti

    2009-01-01

    In this work the main dosimetric characteristics of the Fricke Xylenol Gel (FXG) solution were established for further application in the measurement of dose distribution of clinical electron fields. The dose-response curves of the FXG in a neutron field were also evaluated for the research in Boron Neutron Capture Therapy (BNCT) and industrial electron fields. The standard reading technique was the spectrophotometric. For the clinical field, the intra and inter-batch reproducibility are better than 1.4% and 5.1 %, respectively, the response presents a linear behavior for doses ranging from 0.2 to 40 Gy independently of the energy and the dose rate in the studied ranges. Due to the effects of the FXG natural oxidation, the optimum elapsed time between FXG preparation and irradiation was established as 24h period and the behavior of the dose-response curve of the FXG using the variation in the absorbance relative to the non-irradiated dosimeter as a basis during the whole studied period were not altered. The dose-response to the industrial electron beam presented an exponential decreasing behavior and the neutron beam for research in BNCT presented a linear behavior for the complete studied dose range. According to the obtained results for the different types of radiation studied for the FXG, there was no change in the position of the characteristic bands of the absorption spectrum due to the interaction of these radiation types. Additional tests were performed to determine the digital photographic imaging of FXG analyses viability and the application of FXG dosimetry on intracavitary brachytherapy. The good performance of the FXG dosimeter in the tests that were carried out indicates that this dosimeter may be applied to the tri-dimensional dose evaluation in radiotherapic treatments using electrons and neutron beams. (author)

  8. Texture investigation in aluminium and iron - silicon samples by neutron diffraction technique

    International Nuclear Information System (INIS)

    Pugliese, R.; Yamasaki, J.M.

    1988-09-01

    By means of the neutron diffraction technique the texture of 5% and 98% rolled-aluminium and of iron-silicon steel used in the core of electric transformers, have been determined. The measurements were performed by using a neutron diffractometer installed at the IEA-R1 Nuclear Research Reactor, in the Beam-Hole n 0 . 6. To avoid corrections such as neutron absorption and sample luminosity the geometric form of the samples were approximated to spheric or octagonal prism, and its dimensions do not exceed that of the neutron beam. The texture of the samples were analysed with the help of a computer programme that analyses the intensity of the diffracted neutron beam and plot the pole figures. (author) [pt

  9. Technique investigation on large area neutron scintillation detector array

    International Nuclear Information System (INIS)

    Chen Jiabin

    2006-12-01

    The detailed project for developing Large Area Neutron Scintillation Detector Array (LaNSA) to be used for measuring fusion fuel area density on Shenguang III prototype is presented, including experimental principle, detector working principle, electronics system design and the needs for target chamber etc. The detailed parameters for parts are given and the main causes affecting the system function are analyzed. The realization path is introduced. (authors)

  10. PELAN - a transportable, neutron-based UXO identification technique

    International Nuclear Information System (INIS)

    Vourvopoulos, G.

    1998-01-01

    An elemental characterization method is used to differentiate between inert projectiles and UXO's. This method identifies in a non-intrusive, nondestructive manner, the elemental composition of the projectile contents. Most major and minor chemical elements within the interrogated object (hydrogen, carbon, nitrogen, oxygen, fluorine, phosphorus, chlorine, arsenic, etc.) are identified and quantified. The method is based on PELAN - Pulsed Elemental Analysis with Neutrons. PELAN uses pulsed neutrons produced from a compact, sealed tube neutron generator. Using an automatic analysis computer program, the quantities of each major and minor chemical element are determined. A decision-making tree identifies the object by comparing its elemental composition with stored elemental composition libraries of substances that could be contained within the projectile. In a series of blind tests, PELAN was able to identify without failure, the contents of each shell placed in front of it. The PELAN probe does not need to be in contact with the interrogated projectile. If the object is buried, the interrogation can take place in situ provided the probe can be inserted a few centimeters from the object's surface. (author)

  11. Bone structure investigation using X-ray and neutron radiography techniques

    International Nuclear Information System (INIS)

    Kamali Moghaddam, K.; Taheri, T.; Ayubian, M.

    2008-01-01

    In this paper we report a study of the periodic variation of bone tissue humidity immediately after death using both neutron and X-ray radiography techniques. After death, bone tissue experiences sequential change over time. This change consists of organic and inorganic phase variations of the bone structure, as well as gradual reduction of the bone's water content. These variations are investigated by periodically imaging dead bone using X-ray and neutron radiography. Chemical separation techniques such as calcification and decalcification were used to separate the organic and inorganic phases of the bone. Comparison between X-ray and neutron radiographs of bone following phase separation can be potentially used to investigate the bone disease or to determine a cause of death. In our experiments, we use adult rat femur bones, and the interpretations of these results are presented based on our understanding of bone structure and images produced by neutron and X-ray photon interactions

  12. Bone structure investigation using X-ray and neutron radiography techniques

    Energy Technology Data Exchange (ETDEWEB)

    Kamali Moghaddam, K. [Nuclear Research Center (NRC), Atomic Energy Organization of Iran (AEOI), P.O. Box 11365-8486, Tehran (Iran, Islamic Republic of)], E-mail: kkamali@aeoi.org.ir; Taheri, T.; Ayubian, M. [Nuclear Research Center (NRC), Atomic Energy Organization of Iran (AEOI), P.O. Box 11365-8486, Tehran (Iran, Islamic Republic of)

    2008-01-15

    In this paper we report a study of the periodic variation of bone tissue humidity immediately after death using both neutron and X-ray radiography techniques. After death, bone tissue experiences sequential change over time. This change consists of organic and inorganic phase variations of the bone structure, as well as gradual reduction of the bone's water content. These variations are investigated by periodically imaging dead bone using X-ray and neutron radiography. Chemical separation techniques such as calcification and decalcification were used to separate the organic and inorganic phases of the bone. Comparison between X-ray and neutron radiographs of bone following phase separation can be potentially used to investigate the bone disease or to determine a cause of death. In our experiments, we use adult rat femur bones, and the interpretations of these results are presented based on our understanding of bone structure and images produced by neutron and X-ray photon interactions.

  13. ICF implosion hotspot ion temperature diagnostic techniques based on neutron time-of-flight method

    International Nuclear Information System (INIS)

    Tang Qi; Song Zifeng; Chen Jiabin; Zhan Xiayu

    2013-01-01

    Ion temperature of implosion hotspot is a very important parameter for inertial confinement fusion. It reflects the energy level of the hotspot, and it is very sensitive to implosion symmetry and implosion speed. ICF implosion hotspot ion temperature diagnostic techniques based on neutron time-of-flight method were described. A neutron TOF spectrometer was developed using a ultrafast plastic scintillator as the neutron detector. Time response of the spectrometer has 1.1 ns FWHM and 0.5 ns rising time. TOF spectrum resolving method based on deconvolution and low pass filter was illuminated. Implosion hotspot ion temperature in low neutron yield and low ion temperature condition at Shenguang-Ⅲ facility was acquired using the diagnostic techniques. (authors)

  14. Implementation of the neutron noise technique under the UBERA-6 project

    International Nuclear Information System (INIS)

    Gomez, Angel; Bellino, Pablo A.

    2009-01-01

    Using the neutron noise technique, kinetics parameters estimations and power calibration were performed in the new core of the RA-6 reactor. These activities were carried on under the nuclear start-up of the UBERA-6 project, which consist in the change of core and power increase of the reactor. In a first stage, in joint with the power estimation, the decay constant of the prompt neutrons (α c ) was estimated. Its value was found to agree with the calculations obtained from neutron codes. Lately, in the high power stage, estimators of the calibration factors for the 16 N detection device were obtained. A thorough analysis of the linearity of the instrumentation used was done, and an alternative methodology was applied in order to estimate the aforementioned factor. The calibration factor obtained by the neutron noise technique was in agreement with the one obtained by thermal balance. (author)

  15. Determination of U-235 quantity in fresh fuel elements by neutron coincidence collar technique

    International Nuclear Information System (INIS)

    Almeida, M.C.M. de; Almeida, S.G. de; Marzo, M.A.S.; Moita, L.P.M.

    1990-01-01

    The U-235 quantity per lenght of fresh fuel assemblies of the Angra-I first recharge was determined by Neutron Coincidence Collar technique (N.C.C.). This technique is well-founded in fresh fuel assemblies activation by thermal neutrons from AmLi source to generate U-235 fission neutrons. These neutrons are detected by coincidence method in polyethylene structure where 18 He-3 detectors were placed. The coincidence counting results, in active mode (AmLi), showed 0,7% to standard deviation and equal to 1,49% to mass in 1000s of counting. The accuracies of different calibration methods were evaluated and compared. The results showed that the operator declared values are consistent. This evaluation was part of technical-exchange program between Safeguards Laboratory from C.N.E.N. and Los Alamos National Lab., United States. (author)

  16. Development of in-situ laser based cutting technique for shock absorber rear nut in pressurized heavy water reactors. CP-2.1

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Jain, R.K.; Upadhyaya, B.N.; Choubey, Ambar; Agrawal, D.K.; Oak, S.M.

    2007-01-01

    We have developed a laser based cutting technique for shock absorber rear nuts in pressurized heavy water reactors (PHWRs). This technique has been successfully used for in-situ laser cutting at RAPS-3 reactor. The technique consists of a motorized compact fixture, which holds a fiber optic beam delivery cutting nozzle and can be operated remotely

  17. Development and mastering of production of dysprosium hafnate as absorbing material for control rods of promising thermal neutron reactors

    International Nuclear Information System (INIS)

    Zakharov, A.V.; Risovany, V.D.; Muraleva, E.M.; Sokolov, V.F.

    2011-01-01

    The main advantages of dysprosium hafnate as an absorbing material for LWR control rods are the following: -) unlimited radiation resistance; - two absorbing components, Dy and Hf, increasing physical efficiency of the material compared to Dy 2 O 3 -TiO 2 and alloy 80% Ag - 15% In - 5% Cd; -) variability of physical efficiency by changing a composition, but maintaining other performance characteristics of the material; -) high process-ability due to the absence of phase transients and single-phase structure (solid solution); -) production of high density pellets. Lab-scale mastering of dysprosium hafnate pellets production showed a possibility of material synthesis using a solid-phase method, as well as of dysprosium hafnate pellets production by cold pressing and subsequent sintering. Within a whole range of examined compositions (23 mol% - 75 mol% Dy 2 O 3 ), a single-phase material with a highly radiation resistant fluorite-like structure was produced. Experiments on cold pressing and sintering of pellets confirmed a possibility of producing high quality dysprosium hafnate pellets from synthesized powder. A pilot batch of dysprosium hafnate pellets with standard sizes was produced. The standard sizes corresponded to the absorbing elements of the WWER-1000 control rods and met the main requirements to the absorbing element columns. The pilot batch size was approximately 6 kg. Acceptance testing of the pilot batch of dysprosium hafnate pellets was conducted, fulfillment of the requirements of technical conditions was checked and preirradiation properties of the pellets were examined. High quality of the produced pellets was confirmed, thus, demonstrating a real possibility of producing large batches of the dysprosium hafnate pellets. The next step is the production of test absorbing elements and cluster assemblies for the WWER-1000 control rods with their further installation for pilot operation at one of the Russian nuclear power plants

  18. Radio-analysis of hydrogenous material using neutron back-scattering technique

    International Nuclear Information System (INIS)

    Holly, Wiam Ahmed Alteghany

    2014-10-01

    In this work, we have explored the possibility of using neutron back-scattering technique in performing radio analysis for samples of hydrogenous materials such as explosives, drugs, crude oil and water, looking for different signals that may be used to discriminate these samples. Monte Carlo simulations were carried out to model the detection system and select the optimal geometry as well. The results were determined in terms of the energy spectra of the back-scattered neutrons.(Author)

  19. Detection and identification of explosives and illicit drugs using neutron based techniques

    International Nuclear Information System (INIS)

    Papp, A.; Csikai, J.; Debrecen University,

    2011-01-01

    Some methods developed in collaboration between the ATOMKI and IEP for bulk hydrogen analysis and for the detection and identification of illicit drugs are presented. Advantages and limitations of neutron techniques (reflection, transmission, elastic and inelastic scatterings, leakage spectra and angular yields of Be(d,n), Pu-Be, D-D, D-T and 252 Cf neutrons transmitted from thick samples, effects of hidden materials) are discussed. (author)

  20. Neutron scattering techniques for structural studies of amorphous solids

    International Nuclear Information System (INIS)

    Wright, A.C.; Yarker, C.A.; Johnson, P.A.V.; Wedgwood, F.A.

    1976-01-01

    Preliminary data are presented for a series of glasses using some of the special features of neutron diffraction. In vitreous K 2 O x TiO 2 x 2SiO 2 each Ti atom is on average surrounded by 5.2 oxygen atoms at a distance of 1.95 A, whereas the structure of V 2 O 5 x TeO 2 is dominated by distorted TeO 6 octahedra. Features in the magnetic correlation function for vitreous O x 79Fe 2 O 3 x P 2 O 5 occur at the Fe,Fe distance found in crystalline FePO 4 . (orig.) [de

  1. Neutron noise measurement technique in a coupled reactor

    International Nuclear Information System (INIS)

    Genoud, J.P.

    1976-01-01

    Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

  2. Investigation of Non-Vacuum Deposition Techniques in Fabrication of Chalcogenide-Based Solar Cell Absorbers

    KAUST Repository

    Alsaggaf, Ahmed

    2015-07-01

    The environmental challenges are increasing, and so is the need for renewable energy. For photovoltaic applications, thin film Cu(In,Ga)(S,Se)2 (CIGS) and CuIn(S,Se)2 (CIS) solar cells are attractive with conversion efficiencies of more than 20%. However, the high-efficiency cells are fabricated using vacuum technologies such as sputtering or thermal co-evaporation, which are very costly and unfeasible at industrial level. The fabrication involves the uses of highly toxic gases such as H2Se, adding complexity to the fabrication process. The work described here focused on non-vacuum deposition methods such as printing. Special attention has been given to printing designed in a moving Roll-to-Roll (R2R) fashion. The results show potential of such technology to replace the vacuum processes. Conversion efficiencies for such non-vacuum deposition of Cu(In,Ga)(S,Se)2 solar cells have exceeded 15% using hazardous chemicals such as hydrazine, which is unsuitable for industrial scale up. In an effort to simplify the process, non-toxic suspensions of Cu(In,Ga)S2 molecular-based precursors achieved efficiencies of ~7-15%. Attempts to further simplify the selenization step, deposition of CuIn(S,Se)2 particulate solutions without the Ga doping and non-toxic suspensions of Cu(In,Ga)Se2 quaternary precursors achieved efficiencies (~1-8%). The contribution of this research was to provide a new method to monitor printed structures through spectral-domain optical coherence tomography SD-OCT in a moving fashion simulating R2R process design at speeds up to 1.05 m/min. The research clarified morphological and compositional impacts of Nd:YAG laser heat-treatment on Cu(In,Ga)Se2 absorber layer to simplify the annealing step in non-vacuum environment compatible to R2R. Finally, the research further simplified development methods for CIGS solar cells based on suspensions of quaternary Cu(In,Ga)Se2 precursors and ternary CuInS2 precursors. The methods consisted of post deposition reactive

  3. Visualization and measurement of fluid phenomena using neutron radiography techniques

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Hibiki, Takashi; Fujine, Shigenori; Yoneda, Kenji; Kanda, Keiji; Nishihara, Hideaki; Tsuruno, Akira; Matsubayashi, Masahito; Sobajima, Makoto; Ohtomo, Shoichi.

    1993-01-01

    This paper presents some of the results from recent work performed on the application of neutron radiography to visualization and measurement of fluid phenomena at the Research Reactor Institute of Kyoto University. Experiments have been performed on the following subjects with use of the NR systems at the Japan Research Reactor 3 and the Nuclear Safety Research Reactor of the Japan Atomic Energy Research Institute as well as the Kyoto University Research Reactor: air-water flow in rectangular ducts with 1.0 and 2.4 mm gaps, air-water flow and steam-water flow in a round tube with 4.0 mm inner diameter. The void fraction was measured by processing the images taken by the neutron radiography. The effect of several corrections in image processing was also discussed previously. It was shown that the proposed method could be useful in observing the flow regimes and measuring the void fraction of gas-liquid two-phase flow in narrow channels. (author)

  4. Study on manufacturing technique of synthetic shock-absorbers for underground disposal of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Onodera, Yoshio; Hayashi, Hiromichi; Ebina, Takeo; Nagase, Takako; Torii, Kazuo

    1997-01-01

    On the cse of underground disposal of high level radioactive wastes, natural bentonite is planned to be used for artificial barrier shock-absorber. This is due to expectation of sealing water or adsorbing nuclear materials using swelling and ion-exchanging capacities of smectite, which is a main component of bentonite. In this study, some swelling laminar compounds with various compositions and structures are synthesized to investigate their water sealing and nuclear adsorbing properties. And, according to their results, an optimum material is selected to develop its economic manufacturing method and investigate its alternative possibility for natural bentonite. From such reason, following two titled studies have been executed; 1) Synthesis of the swelling laminar compounds, and 2) Development of manufacturing technique of artificial shock-absorber. In 1995 FY, 1) Detail investigation on synthetic condition of double octahedral type smectite and 2) modeling of the smectite and stability of same type displacement for base of inducing the computer simulation for estimating creation process of optimum materials, were conducted. (G.K.)

  5. Assessment of erbium as candidate burnable absorber for future PWR operaning cycles: A neutronic and fabrication study

    International Nuclear Information System (INIS)

    Asou, M.; Dehaudt, P.; Porta, J.

    1995-01-01

    Erbium begins to play a role in the control of PWR core reactivity. Generally speaking, burnable absorbers were only used to establish fresh core equilibrium. In France, since the possibility of extending irradiation cycles by 12 to 18 months, then up to 24 and 30 months, has been envisaged, there is renewed interest in burnable absorbers. The fabrication of PWR pellets has been investigated, providing high density and a good erbium homogeneity. The pellets characteristics were consistent with the specifications of PWR fuel. However, with the present process, the grain size remains small. Studies in progress now shows that erbium is not only a valuable alternative to gadolinium, for long fuel cycles (≥18 months) but also a new fuel concept. (orig.)

  6. 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

  7. DOSE-Analyzer. A computer program with graphical user interface to analyze absorbed dose inside a body of mouse and human upon external neutron exposure

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Shigemori, Yuji; Sakamoto, Kensaku

    2010-06-01

    DOSE-Analyzer is a computer program to retrieve the dose information from a database and generate a graph through a graphical user interface (GUI). The database is constructed for absorbed dose, fluence, and energy distribution inside a body of mouse and human exposed upon external neutrons, which is calculated by our developed Monte-Carlo simulation method using voxel-based phantom and particle transport code PHITS. The input configurations of irradiation geometry, subject, and energy are set by GUI. The results are tabulated at particle types, i.e. electron, proton, deuteron, triton, and alpha particle, and target organs on a data sheet of Microsoft Office Excel TM . Simple analysis to compare the output values for two subjects is also performed on DOSE-Analyzer. This report is a user manual of DOSE-Analyzer. (author)

  8. Measurement of detector neutron energy response using time-of-flight techniques

    International Nuclear Information System (INIS)

    Janee, H.S.

    1973-09-01

    The feasibility of using time-of-flight techniques at the EG and G/AEC linear accelerator for measuring the neutron response of relatively sensitive detectors over the energy range 0.5 to 14 MeV has been demonstrated. The measurement technique is described in detail as are the results of neutron spectrum measurements from beryllium and uranium photoneutron targets. The sensitivity of a fluor photomultiplier LASL detector with a 2- by 1-inch NE-111 scintillator was determined with the two targets, and agreement in the region of overlap was very good. (U.S.)

  9. Detection of drugs and explosives using neutron computerized tomography and artificial intelligence techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, F.J.O. [Instituto de Engenharia Nuclear, Cidade Universitaria, Rio de Janeiro, CEP 21945-970, Caixa Postal 68550 (Brazil)], E-mail: fferreira@ien.gov.br; Crispim, V.R.; Silva, A.X. [DNC/Poli, PEN COPPE CT, UFRJ Universidade Federal do Rio de Janeiro, CEP 21941-972, Caixa Postal 68509, Rio de Janeiro (Brazil)

    2010-06-15

    In this study the development of a methodology to detect illicit drugs and plastic explosives is described with the objective of being applied in the realm of public security. For this end, non-destructive assay with neutrons was used and the technique applied was the real time neutron radiography together with computerized tomography. The system is endowed with automatic responses based upon the application of an artificial intelligence technique. In previous tests using real samples, the system proved capable of identifying 97% of the inspected materials.

  10. Detection of drugs and explosives using neutron computerized tomography and artificial intelligence techniques

    International Nuclear Information System (INIS)

    Ferreira, F.J.O.; Crispim, V.R.; Silva, A.X.

    2010-01-01

    In this study the development of a methodology to detect illicit drugs and plastic explosives is described with the objective of being applied in the realm of public security. For this end, non-destructive assay with neutrons was used and the technique applied was the real time neutron radiography together with computerized tomography. The system is endowed with automatic responses based upon the application of an artificial intelligence technique. In previous tests using real samples, the system proved capable of identifying 97% of the inspected materials.

  11. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)

    2011-04-01

    The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  12. Radiochemical separation and their application to neutron activation analysis technique

    International Nuclear Information System (INIS)

    Turel, Z.R.

    2013-01-01

    The present paper discusses the development of some new, rapid and selective method for the radiochemical separation and estimation of elements such as, Co(II) 2-3 , Ir(III) 4 , Au(III) 5 , Pt(IV), Pd(II), Os(IV) 6 , Cu(II), Ag(I), Mo(VI), Ni(II), Zn(II), Cd(II), Hg(II), Cs(I), Sb(III), La(III), Sc(III) etc. using various reagents. Various parameters such as pH, time of equilibrium, effect of anions and cations, effect of reagent etc. has been determined employing tracers of the elements under consideration and will be discussed. The method is made highly selective by the use of appropriate masking agent. The stoichiometry of metal reagent is determined by the substoichiometric method. Some examples of multielemental radiochemical separation methods thus developed which have been applied in determining the elements by radiochemical thermal neutron activation analysis will be presented and discussed. The implications of the results on the reference system will also be accounted. Statistical evaluation with reference to accuracy, precision and sensitivity will also be presented

  13. Wear studies of engine components using neutron activation techniques

    International Nuclear Information System (INIS)

    Banados Perez, H.E.; Carvalho, G.; Daltro, T.F.L.

    1984-01-01

    The results obtained in a series of tests for determining the wearing rate of some diesel engine components are reported. The pieces investigated were the needles of fuel injection nozzles, that were previously irradiated with a 10 13 nv in the IEA-R1 nuclear reactor, and the wearing rate was established for different types of fuels. Total wear was calculated by measuring the specific activity of 51 Cr present in the fuel and originated by metal particles worn from the needle. Wear were performed using a device that simulated the actual working conditions of the injection nozzles. The system was run during 350 hours and, along that period, 36 fuel samples of 10 ml each, were collected and analysed for cumulative wear calculation. A metal concentration as low as 10- 6 g in 10 ml of fuel sample could be measured by this method. At present time this procedure is being applied for measuring the wear-rate of other nozzle parts, using localized neutron activation techiques. (Author) [pt

  14. Applications of neutron activation analysis technique in the IPR-R1 research reactor

    International Nuclear Information System (INIS)

    Sabino, C.V.S.; Mansur, N.

    1986-01-01

    A review is made of the neutron activation analysis technique used in the IPR-R1 reactor of the Centro de Desenvolvimento da Tecnologia Nuclear - NUCLEBRAS. Some characteristics of the method are described, types of samples and elements analyzed are also mentioned. (Author) [pt

  15. A cement channel-detection technique using the pulsed-neutron log

    International Nuclear Information System (INIS)

    Myers, G.D.

    1991-01-01

    A channel-detection technique has been developed using boron solutions and pulsed-neutron logging (PNL) tools. This technique relies on the extremely high-neutron-absorption cross section that boron exhibits relative to other common elements, including chlorine. The PNL tool is used to detect movement of a boron solution in a log-inject-log procedure. The technique has identified channels in such difficult applications as logging through two strings of pipe and in highly deviated wellbores. Logging procedures are simple and cement channels can be readily identified. The boron solutions are relatively inexpensive, safe to handle, and nonradioactive. Additional PNL information for reservoir performance evaluation is collected simultaneously during channel-detection logging. This paper describes the theory, development, field application, and limitations of this channel-detection logging technique

  16. Fast neutron detection using a new pulse shape discrimination technique: Charge sensitive integration

    International Nuclear Information System (INIS)

    Zucker, M.; Tsoupas, N.; Karwowski, H.; Castaneda, C.; Nimnual, S.; Porter, R.; Ward, T.

    1988-01-01

    A new electronic technique that depends on charge sensitive integration (CSI) has been developed and tested using a CAMAC based pulse shape discrimination system. Neutrons are well separated from γ-ray signals in the 0.1-100 MeV energy range. The new method was compared with the old zero-crossing time-to-amplitude differentiating technique and was found to be comparable in count rate and superior in noise suppression

  17. Solar powered irrigation management using neutron scattering technique

    International Nuclear Information System (INIS)

    Hegazi, A.M.A

    2010-01-01

    designed (metal frame, detecting sensors and DC motor with gear-box) and used for this purpose. 28% more power output was gained.3.Matching system components with DC motor obtained from local markets to meet PV generator performance. Performance curves of both PV generator and DC motor was plotted.4.Setting up a pumping system matches the PV output, pump efficiency was calculated (30%) at 1100 rpm.5.Hourly pump discharge and reference evapotranspiration (ET o ) were estimated using 10 years meteorological data averages (Meteo-Norm Software), years (1995-2005). Analysis showed that the strongest correlation was between the solar radiation and ET o , while solar radiation with other parameters affecting ET o such as temperature, RH% were not that strong. Correlations between both hourly solar radiation and air temperature were positive and highly significant. Calculations were made to determine area to be cultivated using the pumping system under study, to meet ET o . Results ranged from more than 4 feddan at winter time to 2.3 feddan in summer season. 6. Determination of actual evapotranspiration (ET o ) using neutron moisture meter for the evaluation of estimated (ET o ). Actual evapotranspiration (ET a ) values were calculated within the wet area around emitter (location of plant) in five sites (S1, S2, S3, S4 and S5). Average ET a values for three replicates for the five sites in developing stage of squash plant were calculated.

  18. Practical adjoint Monte Carlo technique for fixed-source and eigenfunction neutron transport problems

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1981-01-01

    An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs

  19. Improvements of neutron activation techniques for the determination of fissile material concentrations

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.

    1987-01-01

    Certain experimental improvements, as variable sample size and irradiation position, automation and flexibility in radiation detection, broaden the measurable concentration range, increase the possible rate and accuracy of analysis and enlarge the application range of home-made nuclear analyzer for fissile material analysis by delayed fission neutron counting and for short-lived multielement analysis by neutron activation gamma-ray spectrometry. Intercomparisons of results by various methods and laboratories show the need for regular checks of techniques to ensure reliable measurements. (author)

  20. Investigation of an egyptian phosphate ore sample by neutron activation analysis technique

    International Nuclear Information System (INIS)

    Eissa, E.A.; Aly, R.A.; Rofail, N.B.; Hassan, A.M.

    1995-01-01

    A domestic phosphate ore sample has been analysed by means of prompt and delayed gamma-ray spectrometry following the activation by thermal neutron capture technique. The rabbit pneumatic transfer system (RPTS), long irradiation facility and two Pu/Be (2,5 Ci each) neutron sources set-Pu for prompt (n,gamma) were applied. The high purity germanium (HPGe) gamma-ray spectrometer with a personal computer analyzer (PCA) system were used for spectrum measurements. Programmes on the VAX computer were utilized for estimating the elemental concentrations of 22 out of 36 elements identified in this work. 2 tabs

  1. Analysis of boron utilization in sample preparation for microorganisms detection by neutron radiography technique

    International Nuclear Information System (INIS)

    Wacha, Reinaldo; Crispim, Verginia R.

    2000-01-01

    The neutron radiography technique applied to the microorganisms detection is the study of a new and faster alternative for diagnosis of infectious means. This work presents the parameters and the effects involved in the use of the boron as a conversion agent, that convert neutrons in a particles, capable ones of generating latent tracks in a solid state nuclear tracks detector, CR-39. The collected samples are doped with the boron by the incubation method, propitiating an interaction microorganisms/boron, that will guarantee the identification of the images of those microorganisms, through your morphology. (author)

  2. Dynamics of polymers in elongational flow studied by the neutron spin-echo technique

    International Nuclear Information System (INIS)

    Rheinstaedter, Maikel C.; Sattler, Rainer; Haeussler, Wolfgang; Wagner, Christian

    2010-01-01

    The nanoscale fluctuation dynamics of semidilute high molecular weight polymer solutions of polyethylenoxide (PEO) in D 2 O under non-equilibrium flow conditions were studied by the neutron spin-echo technique. The sample cell was in contraction flow geometry and provided a pressure driven flow with a high elongational component that stretched the polymers most efficiently. Neutron scattering experiments in dilute polymer solutions are challenging because of the low polymer concentration and corresponding small quasi-elastic signals. A relaxation process with relaxation times of about 10 ps was observed, which shows anisotropic dynamics with applied flow.

  3. Higher order Cambell techniques for neutron flux measurement. Pt. 1

    International Nuclear Information System (INIS)

    Lux, I.; Baranyai, A.

    1982-01-01

    An exact mathematical description of arbitrary high order Campbell techniques for measuring particle fluxes is given. The nth order Campbell technique assumes the measurement of the moments of the outcoming voltage up to the nth one. A simple relation is derived among the various moments of the total measured voltage and of the detector signal caused by one incident particle. It is proven that in the monoparticle case combination of the measured moments up to the order n provides an expression proportional to the particle flux and to the nth moment of the detector signal. Generalization to several different particles is given and it is shown that if the flux of the particle causing the largest detector signal is measured with a relative error epsilon in the dc method and the error is due to the signals of other particles, then in the nth order campbelling the error will be of order epsilonsup(n). The effect of a random background on the measured voltage is also investigated and it is established that the nth order campbelling supresses the noise according to the nth power of the relative amplitude of the noise to the signal. The results concerning constant fluxes are generalized to time dependent particle fluxes and a method assuming a Fourier transform of the measured quantities is proposed for their determination. (orig.)

  4. Implementation of the neutron noise technique for subcritical reactors using a new data acquisition system

    International Nuclear Information System (INIS)

    Bellino, Pablo A.; Gomez, Angel

    2009-01-01

    A new data acquisition system was designed and programmed for nuclear kinetics parameter estimations in subcritical reactors. The system allows using any of the neutron noise techniques, since it could store the whole information available in the neutron detection system. The α Rossi, α Feynman and spectral analysis methods were performed in order to estimate the prompt neutron decay constant (and hence the reactivity). The measurements were done in the nuclear research reactor RA-1, where introducing the control rods, different reactivity levels where reached (until -7 dollars). With the three methods used, agreement was found between the estimations and the reference reactivities in each level, even when the detector efficiency was low. All the measurements were performed with a high gamma flux, although the results were found to be satisfactory. (author)

  5. Technique of neutron-induced (fission-track) autoradiography with histological detail

    International Nuclear Information System (INIS)

    Smith, J.M.; Taylor, G.N.; Jee, W.S.S.

    1980-01-01

    The primary advantage of neutron-induced or fission-track autoradiography compared with conventional autoradiography is that for low concentrations of fissile nuclides prohibitively long exposure times may be avoided. However, it is difficult to produce imaging of biological structures on the neutron-induced autoradiograph which would allow localization of the nuclide histologically. The technique presented circumvents this difficulty using a thin polycarbonate film applied to the histologically stained tissue section mounted on a quartz substrate. After irradiation of the tissue section with an appropriate thermal neutron flux, the fission fragment tracks are revealed by etching the film with KOH. The tracks, superimposed on the stained tissue, may be observed under the light microscope in the same manner as for conventional nuclear emulsion autoradiography

  6. Development of neutron interrogation techniques for detection of hazardous substances in containers port

    International Nuclear Information System (INIS)

    D’Amico, N. M. B; Mayer, R.E; Tartaglione, A.

    2013-01-01

    This work is aimed at contributing to the effort of nations seeking to control international borders movement of dangerous chemical substances and nuclear material, in accordance with a multitude of agreements signed to that purpose. At this stage, we try to identify the signature of pure substances: chlorine (Cl), nitrogen (N), chromium (Cr), mercury (Hg), cadmium (Cd), uranium (U) y arsenic (As) and, later, to detect their presence in simulated large cargo containers. The technique employed in previous and in current work, consists in the detection of prompt and early decay gammas induced by incident thermal neutrons or fast neutrons thermalized in the cargo array. Uranium has also been detected through the counting of fast neutrons originated in induced fissions. (author)

  7. Study of heterogeneous multiplying and non-multiplying media by the neutron pulsed source technique

    International Nuclear Information System (INIS)

    Deniz, V.

    1969-01-01

    The pulsed neutron technique consists essentially in sending in the medium to be studied a short neutron pulse and in determining the asymptotic decay constant of the generated population. The variation of the decay constant as a function of the size of the medium allows the medium characteristics to be defined. This technique has been largely developed these last years and has been applied as well to moderator as to multiplying media, in most cases homogeneous ones. We considered of interest of apply this technique to lattices, to see if useful informations could be collected for lattice calculations. We present here a general theoretical study of the problem, and results and interpretation of a series of experiments made on graphite lattices. There is a good agreement for non-multiplying media. In the case of multiplying media, it is shown that the age value used until now in graphite lattices calculations is over-estimated by about 10 per cent [fr

  8. Neutron diffraction tomography: a unique, 3D inspection technique for crystals using an intensifier TV system

    International Nuclear Information System (INIS)

    Davidson, J.B.; Case, A.L.

    1978-01-01

    The application of phosphor-intensifier-TV techniques to neutron topography and tomography of crystals is described. The older, analogous x-ray topography using wavelengths approximately 1.5A is widely used for surface inspection. However, the crystal must actually be cut in order to see diffraction anomalies beneath the surface. Because 1.5-A thermal neutrons are highly penetrating, much larger and thicker specimens can be used. Also, since neutrons have magnetic moments, they are diffracted by magnetic structures within crystals. In neutron volume topography, the entire crystal or a large part of it is irradiated, and the images obtained are superimposed reflections from the total volume. In neutron tomography (or section topography), a collimated beam irradiates a slice (0.5 to 10 mm) of the crystal. The diffracted image is a tomogram from this part only. A series of tomograms covering the crystal can be taken as the specimen is translated in steps across the narrow beam. Grains, voids, twinning, and other defects from regions down to 1 mm in size can be observed and isolated. Although at present poorer in resolution than the original neutron and film methods, the TV techniques are much faster and, in some cases, permit real-time viewing. Two camera systems are described: a counting camera having a 150 mm 6 Li-ZnS screen for low-intensity reflections which are integrated in a digital memory, and a 300-mm system using analog image storage. Topographs and tomograms of several crystals ranging in size from 4 mm to 80 mm are shown

  9. Gravitational waves from rotating neutron stars and evaluation of fast chirp transform techniques

    CERN Document Server

    Strohmayer, T E

    2002-01-01

    X-ray observations suggest that neutron stars in low mass x-ray binaries (LMXB) are rotating with frequencies in the range 300-600 Hz. These spin rates are significantly less than the break-up rates for essentially all realistic neutron star equations of state, suggesting that some process may limit the spin frequencies of accreting neutron stars to this range. If the accretion-induced spin up torque is in equilibrium with gravitational radiation losses, these objects could be interesting sources of gravitational waves. I present a brief summary of current measurements of neutron star spins in LMXBs based on the observations of high-Q oscillations during thermonuclear bursts (so-called 'burst oscillations'). Further measurements of neutron star spins will be important in exploring the gravitational radiation hypothesis in more detail. To this end, I also present a study of fast chirp transform (FCT) techniques as described by Jenet and Prince (Prince T A and Jenet F A 2000 Phys. Rev. D 62 122001) in the conte...

  10. Fostering applications of neutron scattering techniques in developing countries: IAEA's role

    Energy Technology Data Exchange (ETDEWEB)

    Paranjpe, Shriniwas K. [Division of Physical and Chemical Sciences, International Atomic Energy Agency, Wagramer Strasse 5, A-1400 Vienna (Austria)]. E-mail: S.K.Paranjpe@iaea.org; Mank, G. [Division of Physical and Chemical Sciences, International Atomic Energy Agency, Wagramer Strasse 5, A-1400 Vienna (Austria); Ramamoorthy, N. [Division of Physical and Chemical Sciences, International Atomic Energy Agency, Wagramer Strasse 5, A-1400 Vienna (Austria)

    2006-11-15

    Over the last 60 years research reactors have played an important role in technological and socio-economical development of mankind. Neutron scattering has been the workhorse for research and development in materials science. Developing countries with moderate flux research reactors have also been involved in using this technique. The reactors and the facilities around them have a large potential for applications, while their under-utilization has been a concern for many member states. The International Atomic Energy Agency (IAEA) has been supporting its member states in the enhancement of utilization of their research reactors. Technical meetings focussing on the area of current interests with potential applications are organized under the project on 'effective utilization of research reactors,' e.g. on residual stress measurement, neutron reflectometry. Coordinated research projects (CRPs) bring together scientists from developed and developing countries, build collaborations, and exchange expertise and technology. The CRPs on research reactor utilization include topics like development of small-angle neutron scattering applications and development of sources and imaging systems for neutron radiography. New CRPs on the measurement of residual stress and accelerator-driven neutron sources will be initiated soon. The results from these meetings of CRPs are published as technical documents of the IAEA that would act as guidelines for capacity building for research reactor managers. This paper will present some of the salient features of IAEA activities in promoting research reactor utilization.

  11. Modeling the tagged-neutron UXO identification technique using the Geant4 toolkit

    International Nuclear Information System (INIS)

    Zhou, Y.; Zhu, X.; Wang, Y.; Mitra, S.

    2012-01-01

    It is proposed to use 14 MeV neutrons tagged by the associated particle neutron time-of-flight technique (APnTOF) to identify the fillers of unexploded ordnances (UXO) by characterizing their carbon, nitrogen and oxygen contents. To facilitate the design and construction of a prototype system, a preliminary simulation model was developed, using the Geant4 toolkit. This work established the toolkit environment for (a) generating tagged neutrons, (b) their transport and interactions within a sample to induce emission and detection of characteristic gamma-rays, and (c) 2D and 3D-image reconstruction of the interrogated object using the neutron and gamma-ray time-of-flight information. Using the modeling, this article demonstrates the novelty of the tagged-neutron approach for extracting useful signals with high signal-to-background discrimination of an object-of-interest from that of its environment. Simulations indicated that an UXO filled with the RDX explosive, hexogen (C 3 H 6 O 6 N 6 ), can be identified to a depth of 20 cm when buried in soil. (author)

  12. A helium-3 proportional counter technique for estimating fast and intermediate neutrons

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

    1976-11-01

    3 He proportional counter was employed to determine the fast and intermediate neutron spectra of wide energy region. The mixed gas ( 3 He, Kr) type counter response and the spectrum unfolding code were prepared and applied to some neutron fields. The counter response calculation was performed by using the Monte Carlo code, paying regards to dealing of the particle range calculation of the mixed gas. An experiment was carried out by using the van de Graaff accelerator to check the response function. The spectrum unfolding code was prepared so that it may have the function of automatic evaluation of the higher energy spectrum's effect to the pulse hight distribution of the lower energy region. The neutron spectra of the various neutron fields were measured and compared with the calculations such as the discrete ordinate Sn calculations. It became clear that the technique developed here can be applied to the practical use in the neutron energy range from about 150 KeV to 5 MeV. (auth.)

  13. A numerical approach to the time dependent neutron flux using the Laplace transform technique

    International Nuclear Information System (INIS)

    El-Demerdash, A; Beynon, T.D.

    1979-01-01

    In this study a time dependent transport problem in which an isotopic neutron source emits a pulse of neutrons into a finite sphere has been solved by a numerical Laplace transform technique. The object has been to investigate the time behaviour of the neutron field in the moderators at times shortly after the neutron source initiation, that is in the nanosecond time period. The basis of the solution is a numercial evaluation of the Laplace transform of the flux in the linear Boltzmann equation with the use of a modified version of a steady state energy multi-group spatially dependent code. The explicit or direct inversion of the Laplace transformed flux is complicated to be solved numerically due to the ill-conditioned matrix obtained. The suggested method of solutions depends on choice of a function that satisfies the physical condition known from the neutron behaviour and that has a Laplace inversion which is analytically amenable. By employing a least square fitting procedure the function is modified in order to minimize the error in the Laplace transformed values and hence in the time dependent solution. This method has been applied satisfactorily in comparison to analytical and experimental results

  14. Non-destructive test of lock actuator component using neutron radiography technique

    International Nuclear Information System (INIS)

    Juliyanti; Setiawan; Sutiarso

    2012-01-01

    Non-destructive test of lock actuator using neutron radiography technique has been done. The lock actuator is a mechanical system which is controlled by central lock module consisting of electronic circuit which drives the lock actuator works accordingly to open and lock the vehicle door. The non-destructive test using neutron radiography is carried out to identify the type of defect is presence by comparing between the broken and the brand new one. The method used to test the lock actuator component is film method (direct method). The result show that the radiography procedure has complied with the ASTM standard for neutron radiography with background density of 2.2, 7 lines and 3 holes was seen in the sensitivity indicator (SI) and the quite good image quality was obtained. In the brand new actuator is seen that isolator part which separated the coils has melted. By this non-destructive test using neutron radiography technique is able to detect in early stage the type of component's defect inside the lock actuator without to dismantle it. (author)

  15. Neutron Backscattered Technique for Quantification of Oil Palm Fruit Oil Content

    International Nuclear Information System (INIS)

    Ismail Mustapha; Samihah Mustaffha; Md Fakarudin Ab Rahman; Roslan Yahya; Lahasen Norman Shah Dahing; Nor Paiza Mohd Hasan; Jaafar Abdullah

    2013-01-01

    Non-destructive and real time method becomes a well-liked method to researchers in the oil palm industry since 2000. This method has the ability to detect oil content in order to increase the production of oil palm for better profit. Hence, this research investigates the potential of neutron source to estimate oil content in palm oil fruit since oil palm contains hydrogen with chemical formula C 55 H 96 O 6 . For this paper, oil palm loose fruit was being used and divided into three groups. These three groups are ripe, under-ripe and bruised fruit. A total of 21 loose fruit for each group were collected from a private plantation in Malaysia. Each sample was scanned using neutron backscattered technique. The higher neutron count, the more hydrogen content, and the more oil content in palm oil fruit. The best correlation result came from the ripe fruits with r 2 =0.98. This research proves that neutron backscattered technique can be used as a non-destructive and real time grading system for palm oil. (author)

  16. Application of the neutron noise analysis technique in nuclear power plants

    International Nuclear Information System (INIS)

    Lescano, Victor H.; Wentzeis, Luis M.

    1999-01-01

    Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

  17. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    International Nuclear Information System (INIS)

    Fanaro, L.C.C.B.

    1984-01-01

    It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

  18. Studies review and exploration purpose of neutron radiography technique in the TRIGA IPR-R1 reactor at CDTN, Brazil

    International Nuclear Information System (INIS)

    Costa, Antonella Lombardi; Amorim, Valter Alves de; Stasiulevicius, Roberto; Rocha, Zildete

    2002-01-01

    Neutron Radiography - NR - consists of obtaining on a sensitive plate, the image produced by neutron flux after crossing an object. Through NR is possible to inspect plastics and explosives materials and organic composition. Is difficult to analyze these materials by the radiography technique. The neutron beam extractor was installed, in the TRIGA IPR-R1 reactor at the CDTN. This work presents preliminaries results of the NR researches in the past at CDTN, which are being retaken. (author)

  19. Use of borosilicate-glass raschig rings as a neutron absorber in solutions of fissile material-ANSI/ANS-8.5-1996

    International Nuclear Information System (INIS)

    Rothe, R.E.; Ketzlach, N.; Finch, D.R.

    1996-01-01

    American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.5 is one of several standards prepared by the ANS Standards Committee to provide guidance to enhance criticality safety in the handling, storage, and processing of fissionable materials. American National Standard ANSI/ANS-8.5-1996 provides this guidance for one type of boron-loaded glass in one type of geometry (cylindrical rings) for use with fissile solutions. Recorded use of such fixed neutron absorbers for criticality control of fissile solutions dates back to 1958, but some less-well-documented applications were recorded as early as the mid-1940's. The first solid efforts to collect recommendations derived from experience and technology were begun in 1965. Over the next 6 yr additional experiments were performed, and supporting data for the proposed standard were gathered. The first standard on this safety matter was issued in 1971. It was reaffirmed in 1979 with only minor changes and a slight expansion of the coverage. The standard was last revised in 1986

  20. The determination of pesticide residues in local vegetables by means of neutron activation technique

    International Nuclear Information System (INIS)

    Mongkolphantha, S.; Karasuddhi, P.; Yamkate, P.; Serichareonsatit, N.

    1975-01-01

    Analytical methods based on neutron activation have been developed for studying pesticides residues of bromine, arsenic and mercury in local vegetables and fruits. The concentration of bromine, arsenic and mercury in samples are enriched prior to neutron irradiations by a technique of dry-ashing and freeze-drying for the determination of arsenic, bromine and mercury respectively. The element bromine is determined instrumentally while arsenic and mercury are determined destructively using a distillation technique. The limit of detection under the conditions used for bromine, arsenic and mercury as obtained are 0.01, 0.001 and 0.0001 microgram respectively. A total of 45 varieties of vegetables and 20 varieties of fruits are analyzed. The results of the investigation and the concentration range in part per million of bromine, arsenic and mercury are also presented

  1. Basic research for developing the quantitative neutron radiography

    International Nuclear Information System (INIS)

    Tamaki, Masayoshi; Ikeda, Yasushi; Ohkubo, Kohei; Tasaka, Kanji; Yoneda, Kenji; Fujine, Shigenori.

    1992-01-01

    This investigation concerns the basic research and development on quantitative neutron radiography by using a honeycomb collimator which reduces the effect due to scattered neutrons in objective matter. On the observation of the hydrogenate materials such as metal hydrides, water and hydrocarbons by neutron radiography, scattered neutrons from these objectives make the quantitativeness of the neutron radiographic image lower grade. In order to improve the quantitativeness of the image, a honeycomb collimator, which is a honeycomb structure of neutron absorbing material, was introduced to the conventional neutron radiography system. By setting the neutron-absorbing honeycomb collimator between objective and imaging system, neutrons scattered in the objective were absorbed by the honeycomb material and attenuated before coming to the imaging system, but neutrons which were transmitted the objective sample without any interaction reached the imaging system and formed the image of the sample. As the image by purely transmitted neutrons is intrinsic due to the neutronic character of the sample, the image data give the quantitative information. In the present experiment, aluminum honeycomb which was coated with boron nitride was prepared and used in order to image the standard stepwise samples for the evaluation of the quantitative grade of the newly proposed neutron radiography method. From the comparison between macroscopic total cross section and the attenuation coefficient of the thermal neutron for aluminum, copper and hydrocarbons, it was confirmed that they were fairly consistent each other. It can be concluded that the newly proposed neutron radiography method using the neutron-absorbing honeycomb collimator for the elimination of the scattered neutrons improves remarkably the quantitativeness of the neutron radiography technique. (author)

  2. Dynamical scaling in polymer solutions investigated by the neutron spin echo technique

    International Nuclear Information System (INIS)

    Richter, D.; Ewen, B.

    1979-01-01

    Chain dynamics in polymer solutions was investigated by means of the recently developed neutron spin echo spectroscopy. - By this technique, it was possible for the first time to verify unambiguously the scaling predictions of the Zimm model in the case of single chain behaviour and to observe the cross over to many chain behaviour. The segmental diffusion of single chains exhibits deviations from a simple exponential law, indicating the importance of memory effects. (orig.) [de

  3. Trace element analysis at the Livermore pool-type reactor using neutron activation techniques

    International Nuclear Information System (INIS)

    Ragaini, R.C.; Ralston, R.; Garvis, D.

    1975-01-01

    The capabilities of trace element analysis at the Livermore Pool-Type Reactor (LPTR) using instrumental neutron activation analysis (INAA) are discussed. A description is given of the technology and the methods employed, including sample preparation, irradiation, and analysis. Applications of the INAA technique in past and current projects are described. A computer program, GAMANAL, has been used for nuclide identification and quantification. (U.S.)

  4. Viability of neutron brachytherapy technique using Cf252 for tumors of salivary glands therapy

    International Nuclear Information System (INIS)

    Andrade, Lidia M.; Campos, Tarcisio P.R.

    2000-01-01

    Gland salivary tumors, although uncommon, present significant characteristics such as distant metastasis, recurrence and average survive of five year for treated patients. Those tumors presents low response to a conventional radiotherapy, photon and electron therapy; thus, this modality is associated with surgery. Fast neutron therapy has been presented better results in controlling the local tumor in comparison with conventional radiotherapy. Brachytherapy with Cf-252 presents an alternative treatment for tumors. Those radiotherapy technique are discussed and analyzed. (author)

  5. Determination of hydrogen content of petroleum products from Tema Oil Refinery using neutron backscatter technique

    International Nuclear Information System (INIS)

    Salifu, A. S.

    2015-01-01

    The hydrogen content of hydrocarbon materials is very important in several areas of industrial process such as mining, vegetable oil extraction and crude oil exploration and refining. A fast and more universal technique based on thermal neutron reflection was employed to determine the total hydrogen contents of petroleum samples from Tema Oil Refinery (TOR) and Crude oil samples from Jubilee field and Nigeria. The experimental set-up consisted of a source-holder housing a 1Ci Am-Be neutron source and a He-3 neutron detector. Two geometrical arrangements were considered and their sensitivities were compared. The set-up was used to measure the excess neutron count in both geometrical considerations and their reflection parameters were calculated as a function of hydrogen content of the samples. Calibration lines were deduced using liquid hydrocarbons containing well-known hydrogen and carbon contents as standards. Two linear equations were generated from the calibration lines and were used to further determine the hydrogen content of thirteen (13) petroleum samples obtained from Quality Control Department of TOR. The total hydrogen contents were found to be in the range of 7.211(hw %) - 15.069 (hw %) for vertical geometry and 7.206 (hw %) - 14.948 (hw %) for horizontal geometry respectively. The results obtained agreed constructively with other results obtained using different methodologies by other studies. The percentage error of the hydrogen contents denoted by (% E) for the various petroleum samples were also obtained and noticed to be within an acceptable range. The neutron backscatter technique was observed as an alternative and more generalized method for quality assurance and standardization in the petroleum industries

  6. Genetic algorithms - A new technique for solving the neutron spectrum unfolding problem

    International Nuclear Information System (INIS)

    Freeman, David W.; Edwards, D. Ray; Bolon, Albert E.

    1999-01-01

    A new technique utilizing genetic algorithms has been applied to the Bonner sphere neutron spectrum unfolding problem. Genetic algorithms are part of a relatively new field of 'evolutionary' solution techniques that mimic living systems with computer-simulated 'chromosome' solutions. Solutions mate and mutate to create better solutions. Several benchmark problems, considered representative of radiation protection environments, have been evaluated using the newly developed UMRGA code which implements the genetic algorithm unfolding technique. The results are compared with results from other well-established unfolding codes. The genetic algorithm technique works remarkably well and produces solutions with relatively high spectral qualities. UMRGA appears to be a superior technique in the absence of a priori data - it does not rely on 'lucky' guesses of input spectra. Calculated personnel doses associated with the unfolded spectra match benchmark values within a few percent

  7. An investigation into the suitability of additive manufacturing techniques for neutron moderator vessels

    International Nuclear Information System (INIS)

    Gallimore, S.

    2016-01-01

    Additive manufacturing (also known as rapid prototyping or 3D printing) techniques are increasing in popularity for several key reasons; greater freedom in possible geometry, reduced time of manufacture and connected to these are potential cost savings. ISIS has begun an investigation into the suitability of the various available techniques for the manufacture of neutron moderator vessels, in order to see if it can exploit these advantages. It is however understood that additive manufacturing is by no means a perfect technique and part of the investigations will be to try and better understand how some of the disadvantages of the technique affect its potential application within the spallation neutron environment. Some of the main disadvantages commonly listed are; the grades of materials available/suitable for the process are limited, virtually no pre-existing material data from radiation environments, lower quality surface finish (directly from the manufacturing process), less familiarity with residual stresses in the material and questions over whether tight tolerances and consistent material thicknesses be achieved? The work has been divided into two streams; one which utilises small samples to evaluate and compare different manufacturing and post-treatment techniques, the other that performs tests on a full-size representative moderator vessel. The complete programme of testing shall include the following tests; fundamental 'neutronic transparency', room temperature vacuum leak test, cold shock (using LN_2) and subsequent room temperature leak test, pressure cycling, a burst test, welding suitability and material data testing. The investigations being conducted at ISIS are very much in the early stages and looking at fairly fundamental questions. Answering these will clearly guide the decision whether is it worth continuing with further investigation and development or if the currently available techniques do not produce materials that are suitable for use as

  8. Burnable absorber coated nuclear fuel

    International Nuclear Information System (INIS)

    Chubb, W.; Radford, K.C.; Parks, B.H.

    1984-01-01

    A nuclear fuel body which is at least partially covered by a burnable neutron absorber layer is provided with a hydrophobic overcoat generally covering the burnable absorber layer and bonded directly to it. In a method for providing a UO 2 fuel pellet with a zirconium diboride burnable poison layer, the fuel body is provided with an intermediate niobium layer. (author)

  9. Comparative study between the PIXE technique and neutron activation analysis for Zinc determination

    International Nuclear Information System (INIS)

    Cruvinel, Paulo Estevao; Crestana, Silvio; Artaxo Netto, Paulo Eduardo

    1997-01-01

    This work presents a comparative study between the PIXE, proton beams and neutron activation analysis (NAA) techniques, for determination of total zinc concentration. Particularly, soil samples from the Pindorama, Instituto Agronomico de Campinas, Sao Paulo State, Brazil, experimental station have been analysed and measuring the zinc contents in μg/g. The results presented good correlation between the mentioned techniques. The PIXE and NAA analyses have been carried out by using the series S, 2.4 MeV proton beams Pelletron accelerator and the IPEN/CNEN-IEA-R1 reactor, both installed at the Sao Paulo - Brazil university

  10. Development, improvement and calibration of neutronic reaction rates measurements: elaboration of a standard techniques basis

    International Nuclear Information System (INIS)

    Hudelot, J.P.

    1998-06-01

    In order to improve and to validate the neutronics calculation schemes, perfecting integral measurements of neutronics parameters is necessary. This thesis focuses on the conception, the improvement and the development of neutronics reaction rates measurements, and aims at building a base of standard techniques. Two subjects are discussed. The first one deals with direct measurements by fission chambers. A short presentation of the different usual techniques is given. Then, those last ones are applied through the example of doubling time measurements on the EOLE facility during the MISTRAL 1 experimental programme. Two calibration devices of fission chambers are developed: a thermal column located in the central part of the MINERVE facility, and a calibration cell using a pulsed high flux neutron generator and based on the discrimination of the energy of the neutrons with a time-of-flight method. This second device will soon allow to measure the mass of fission chambers with a precision of about 1 %. Finally, the necessity of those calibrations will be shown through spectral indices measurements in core MISTRAL 1 (UO 2 ) and MISTRAL 2 (MOX) of the EOLE facility. In each case, the associated calculation schemes, performed using the Monte Carlo MCNP code with the ENDF-BV library, will be validated. Concerning the second one, the goal is to develop a method for measuring the modified conversion ratio of 238 U (defined as the ratio of 238 U capture rate to total fission rate) by gamma-ray spectrometry of fuel rods. Within the framework of the MISTRAL 1 and MISTRAL 2 programmes, the measurement device, the experimental results and the spectrometer calibration are described. Furthermore, the MCNP calculations of neutron self-shielding and gamma self-absorption are validated. It is finally shown that measurement uncertainties are better than 1 %. The extension of this technique to future modified conversion ratio measurements for 242 Pu (on MOX rods) and 232 Th (on

  11. Development, improvement and calibration of neutronic reaction rate measurements: elaboration of a base of standard techniques

    International Nuclear Information System (INIS)

    Hudelot, J.P.

    1998-01-01

    In order to improve and to validate the neutronic calculation schemes, perfecting integral measurements of neutronic parameters is necessary. This thesis focuses on the conception, the improvement and the development of neutronic reaction rates measurements, and aims at building a base of standard techniques. Two subjects are discussed. The first one deals with direct measurements by fission chambers. A short presentation of the different usual techniques is given. Then, those last ones are applied through the example of doubling time measurements on the EOLE facility during the MISTRAL 1 experimental programme. Two calibration devices of fission chambers are developed: a thermal column located in the central part of the MINERVE facility, and a calibration cell using a pulsed high flux neutron generator and based on the discrimination of the energy of the neutrons with a time-of-flight method. This second device will soon allow to measure the mass of fission chambers with a precision of about 1 %. Finally, the necessity of those calibrations will be shown through spectral indices measurements in core MISTRAL 1 (UO 2 ) and MISTRAL 2 (MOX) of the EOLE facility. In each case, the associated calculation schemes, performed using the Monte Carlo MCNP code with the ENDF-BV library, will be validated. Concerning the second one, the goal is to develop a method for measuring the modified conversion ratio of 238 U (defined as the ratio of 238 U capture rate to total fission rate) by gamma-ray spectrometry of fuel rods. Within the framework of the MISTRAL 1 and MISTRAL 2 programmes, the measurement device, the experimental results and the spectrometer calibration are described. Furthermore, the MCNP calculations of neutron self-shielding and gamma self-absorption are validated. It is finally shown that measurement uncertainties are better than 1 %. The extension of this technique to future modified conversion ratio measurements for 242 Pu (on MOX rods) and 232 Th (on Thorium

  12. Characterization of plastic and boron carbide additive manufactured neutron collimators

    Science.gov (United States)

    Stone, M. B.; Siddel, D. H.; Elliott, A. M.; Anderson, D.; Abernathy, D. L.

    2017-12-01

    Additive manufacturing techniques allow for the production of materials with complicated geometries with reduced costs and production time over traditional methods. We have applied this technique to the production of neutron collimators for use in thermal and cold neutron scattering instrumentation directly out of boron carbide. We discuss the design and generation of these collimators. We also provide measurements at neutron scattering beamlines which serve to characterize the performance of these collimators. Additive manufacturing of parts using neutron absorbing material may also find applications in radiography and neutron moderation.

  13. Shock absorber

    International Nuclear Information System (INIS)

    Housman, J.J.

    1978-01-01

    A shock absorber is described for use in a hostile environment at the end of a blind passage for absorbing impact loads. The shock absorber includes at least one element which occupies the passage and which is comprised of a porous brittle material which is substantially non-degradable in the hostile environment. A void volume is provided in the element to enable the element to absorb a predetermined level of energy upon being crushed due to impact loading

  14. Second international comparison on measuring techniques of tritium production rate for fusion neutronics experiments (ICMT-2)

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Maekawa, Hiroshi

    1993-02-01

    An second international comparison on measuring techniques of tritium production rates for fusion neutronics experiments (ICMT-2) has been performed. The purpose is to evaluate the measurement accuracy of tritium production rates in the current measurement techniques. Two 14 MeV neutron source facilities, FNS at JAERI-Japan and LOTUS at EPFL-Switzerland, were used for this purpose. Nine groups out of seven countries participated in this program. A fusion simulated blanket assembly of simple-geometry was served as the test bed at each facility, in which Li-containing samples from the participants were irradiated in an uniform neutron field. The tritium production rates were determined by the participants using their own ways by using the liquid scintillation counting method. Tritiated water sample with unknown but the same concentration was also distributed and its concentration was measured to make a common reference. The standard deviation of measured tritium production rates among participants was about 10 % for both FNS and LOTUS irradiation levels: 4x10 -13 T-atoms/Li-atom and 1.6x10 -12 T-atoms/Li-atom at a sample, respectively. This standard deviation exceeds the expected deviation of 5 % in this program. It is presumed that the deviation of 10 % is caused mainly by the systematic and unknown errors in a process of tritium extraction from the irradiated samples depending on each organization. (author)

  15. Resolution function in deep inelastic neutron scattering using the Foil Cycling Technique

    International Nuclear Information System (INIS)

    Pietropaolo, A.; Andreani, C.; Filabozzi, A.; Pace, E.; Senesi, R.

    2007-01-01

    New perspectives for epithermal neutron spectroscopy are being opened up by the development of the Resonance Detector (RD) and its use on inverse geometry time of flight (TOF) spectrometers at spallation sources. The most recent result is the Foil Cycling Technique (FCT), which has been developed and applied on the VESUVIO spectrometer operating in the RD configuration. This technique has demonstrated its capability to improve the resolution function of the spectrometer and to provide an effective neutron and gamma background subtraction method. This paper reports a detailed analysis of the line shape of the resolution function in Deep Inelastic Neutron Scattering (DINS) measurements on VESUVIO spectrometer, operating in the RD configuration and employing the FCT. The aim is to provide an analytical approximation for the analyzer energy transfer function, an useful tool for data analysis on VESUVIO. Simulated and experimental results of DINS measurements on a lead sample are compared. The line shape analysis shows that the most reliable analytical approximation of the energy transfer function is a sum of a Gaussian and a power of a Lorentzian. A comparison with the Double Difference Method (DDM) is also discussed. It is shown that the energy resolution improvement for the FCT and the DDM is almost the same, while the counting efficiency is a factor of about 1.4 higher for the FCT

  16. Software development of the mechanical vibration monitoring system of the CNA I reactor internals by neutron noise technique

    International Nuclear Information System (INIS)

    Wentzeis, Luis M.; Calvo, Maria D.

    2009-01-01

    The neutron noise analysis technique is an important predictive maintenance tool for early detection of failures such as sensor malfunctions and incipient mechanical problems located in the reactor internals. This technique was applied successfully in Argentina since 1987. The FER-GAEN group dependent of the CNEA developed the measuring system to detect anomalies as early as possible. The magnitude of interest in this analysis is the fluctuating component of the neutron flux known as 'neutron noise'. In order to improve and facilitate the analysis, a new software code was developed for the data acquisition of the neutron noise signals and neutron spectra estimation in the frequency domain. The RMS values related with the internals vibrations are calculated from these spectra and are chronologically displayed, in order to detect any anomalous vibration or incipient detector malfunction as early as possible. (author)

  17. The application of moisture measurement in materials used in ironmaking by neutron moderation techniques

    International Nuclear Information System (INIS)

    Williams, R.B.

    1979-01-01

    Nucleonic moisture gauges differ in design geometry so that for each installation, whether it be a large coke weigh hopper, smaller sinter mix hopper or a conveyor belt, there is a design which would best fit the requirements. The types of gauges currently being used extensively in BSC are insert-probe type gauges and surface-mounted type gauges which make use of the neutron backscatter principle. Another much less commonplace technique is that of fast neutron transmission through material but is has been thoroughly tested in on-site trials for a sinter mix belt situation. These three gauge designs are described and discussed furthermore: practical experience in the British Steel Industry will be described. Some operational as well as sensitivity data are given and calibration curves are shown. (T.G.)

  18. Study of some environmental problem in egypt using neutron activation analysis techniques

    International Nuclear Information System (INIS)

    El-Karim, A.H.M.G.

    2003-01-01

    this thesis deals with the investigation of the possibility of using the new (second) egyptian research reactor (ETRR-2) at Inshas (22 MW) for the neutron activation analysis (ANN) of trace elements, particularly in air dust, collected from cairo and some other cities of egypt. in this concern chapter 1 gives an introduction about the activation methods in general, describing the various techniques used and a comparison of the methods with other instrumental methods of analysis . as a main classification, the neutron activation methods involve prompt γ-ray NAA and delayed γ-ray NAA; cyclic NAA (repeated activation) was also outlined. the methodology of NAA involves the absolute method, the relative method and the mono standard (single comparator) method , which is in between the absolute and relative methods

  19. The VALMONT Program: Accurate experimental techniques to support the neutronics qualification of UMo/AL

    International Nuclear Information System (INIS)

    Hedelot, J.P.; Doederlein, C.; Antony, M.; Girard, J.M.; Laval, V.; Fougeras, P.; Willermoz, G.; Leconte, P.

    2005-01-01

    The VALMONT program aimed at qualifying the HORUS3D (HOrowitz Reactor simulation Unified System) neutronics calculation route that is used for the development of the JHR core, and to verify the correct treatment of UMo/Al (20% enrichment in 235 U) fuel. The program is composed of two parts. The first part was devoted to the measurement by the oscillation technique of the reactivity effect of UAl/Mo fuel with an accuracy around 1% (1□). The second part consisted of gamma-spectroscopy experiments on a dedicated UMo/Al fuel sample in order to characterize, through axial power profiles and modified conversion ratio of 238 U measurements, the production and absorption effects inside the UAl/Mo fuel. The overall excellent agreement between high accuracy experiments and calculations allowed to qualify the HORUS3D neutronics calculation route for UMo/Al fuel. (author)

  20. Methodological aspects and development of techniques for neutron activation analysis of microcomponents in materials of geologic origin

    International Nuclear Information System (INIS)

    Cohen, I.M.

    1982-01-01

    Some aspects of the activation analysis methodology applied to geological samples activated in nuclear reactors were studied, and techniques were developed for the determination of various elements in different types of matrixes, using gamma spectrometry for the measurement of the products. The consideration of the methodological aspects includes the study of the working conditions, the preparation of samples and standards, irradiations, treatment of the irradiated material, radiochemical separation and measurement. Experiments were carried out on reproducibility and errors in relation to the behaviour of the measurement equipment and that of the methods of area calculation (total area, Covell and Wasson), as well as on the effects of geometry variations on the results of the measurements, the RA-3 reactors's flux variations, and the homogeneity of the samples and standards. Also studied were: the selection of the conditions of determination, including the irradiation and decay times; the irradiation with thermal and epithermal neutrons; the measurement with the use of absorbers, and the resolution of complex peaks. Both non-destructive and radiochemical separation techniques were developed for the analysis of 5 types of geological materials. These methods were applied to the following determinations: a) In, Cd, Mn, Ga and Co in blende; b) La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb and Lu in fluorites; c) La, Ca, Eu, Tb, Yb, Se and Th in barites and celestites; d) Cu and Zn in soils. The spectral interferences or those due to nuclear reactions were studied and evaluated by mathematical calculation. (M.E.L.) [es

  1. Software tool for resolution of inverse problems using artificial intelligence techniques: an application in neutron spectrometry

    International Nuclear Information System (INIS)

    Castaneda M, V. H.; Martinez B, M. R.; Solis S, L. O.; Castaneda M, R.; Leon P, A. A.; Hernandez P, C. F.; Espinoza G, J. G.; Ortiz R, J. M.; Vega C, H. R.; Mendez, R.; Gallego, E.; Sousa L, M. A.

    2016-10-01

    The Taguchi methodology has proved to be highly efficient to solve inverse problems, in which the values of some parameters of the model must be obtained from the observed data. There are intrinsic mathematical characteristics that make a problem known as inverse. Inverse problems appear in many branches of science, engineering and mathematics. To solve this type of problem, researches have used different techniques. Recently, the use of techniques based on Artificial Intelligence technology is being explored by researches. This paper presents the use of a software tool based on artificial neural networks of generalized regression in the solution of inverse problems with application in high energy physics, specifically in the solution of the problem of neutron spectrometry. To solve this problem we use a software tool developed in the Mat Lab programming environment, which employs a friendly user interface, intuitive and easy to use for the user. This computational tool solves the inverse problem involved in the reconstruction of the neutron spectrum based on measurements made with a Bonner spheres spectrometric system. Introducing this information, the neural network is able to reconstruct the neutron spectrum with high performance and generalization capability. The tool allows that the end user does not require great training or technical knowledge in development and/or use of software, so it facilitates the use of the program for the resolution of inverse problems that are in several areas of knowledge. The techniques of Artificial Intelligence present singular veracity to solve inverse problems, given the characteristics of artificial neural networks and their network topology, therefore, the tool developed has been very useful, since the results generated by the Artificial Neural Network require few time in comparison to other techniques and are correct results comparing them with the actual data of the experiment. (Author)

  2. 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).

  3. Mechanical shock absorber

    International Nuclear Information System (INIS)

    Vrillon, Bernard.

    1973-01-01

    The mechanical shock absorber described is made of a constant thickness plate pierced with circular holes regularly distributed in such a manner that for all the directions along which the strain is applied during the shock, the same section of the substance forming the plate is achieved. The shock absorber is made in a metal standing up to extensive deformation before breaking, selected from a group comprising mild steels and austenitic stainless steels. This apparatus is used for handling pots of fast neutron reactor fuel elements [fr

  4. Direct measurement of the cross section of neutron-neutron scattering at the YAGUAR reactor. Substantiation of the experiment technique

    International Nuclear Information System (INIS)

    Chernukhin, Yu.G.; Kandiev, Ya.Z.; Lartsev, V.D.; Levakov, B.G.; Modestov, D.G.; Simonenko, V.A.; Streltsov, S.I.; Khmel'nitskij, D.V.

    2006-01-01

    The main stage of experiment for direct measurement of cross section of neutron-neutron scattering σ nn at low energies (E nn determination. It was shown, that for achieving the criterion ε ∼ 4% it will be necessary to have 40-50 pulses of a reactor [ru

  5. A neutron-X-ray, NMR and calorimetric study of glassy Probucol synthesized using containerless techniques

    International Nuclear Information System (INIS)

    Weber, J.K.R.; Benmore, C.J.; Tailor, A.N.; Tumber, S.K.; Neuefeind, J.; Cherry, B.; Yarger, J.L.; Mou, Q.; Weber, W.; Byrn, S.R.

    2013-01-01

    Highlights: • Acoustic levitation was used to make phase-pure glassy forms of pharmaceutical compounds. • Neutrons, X-rays and NMR were used to characterize the glasses. • The glass comprised of slightly distorted molecules packed in a random network. • Potential for new drug synthesis routes is discussed. - Abstract: Acoustic levitation was used to trap 1–3 mm diameter drops of Probucol and other pharmaceutical materials in containerless conditions. Samples were studied in situ using X-ray diffraction and ex situ using neutron diffraction, NMR and DSC techniques. The materials were brought into non-equilibrium states by supersaturating solutions or by supercooling melts. The glass transition and crystallization temperatures of glassy Probucol were 29 ± 1 and 71 ± 1 °C respectively. The glassy form was stable with a shelf life of at least 8 months. A neutron/X-ray difference function of the glass showed that while molecular sub-groups remain rigid, many of the hydrogen correlations observed in the crystal become smeared out in the disordered material. The glass is principally comprised of slightly distorted Form I Probucol molecules with disordered packing rather than large changes in the individual molecular structure. Avoiding surface contact-induced nucleation provided access to highly non-equilibrium phases and enabled synthesis of phase-pure glasses

  6. The Comparison Study of Neutron Activation Analysis and Fission Track Technique for Uranium Determination

    International Nuclear Information System (INIS)

    Sirinuntavid, Alice; Rodthongkom, Chouvana

    2007-08-01

    Full text: Comparison between Neutron Activation Analysis (NAA) and fission track technique for uranium determination in solid samples was studied by use of standard reference materials, i.e., ore, coal fly ash, soil. For NAA, the epithermal neutron was applied for activated irradiation. Then, the 74.5 keV gamma from U-239 or 277.7 keV gamma from Np-239 was measured. For high Uranium content samples, NAA method with 74.5 keV gamma measurement, gave higher precision result than the 277.7 keV gamma measurement method. NAA method with 277.7 keV gamma measurement, gave higher sensitivity and precision result for low Uranium content samples and the uranium contained less than 10 ppm samples. Nevertheless, the latter procedure needed longer time for neutron irradiation and analysis procedure. In comparison the results of Uranium analysis between NAA and fission track, it was found that no significant difference within 95 % of confidence level

  7. A neutron-X-ray, NMR and calorimetric study of glassy Probucol synthesized using containerless techniques

    Energy Technology Data Exchange (ETDEWEB)

    Weber, J.K.R., E-mail: rweber@anl.gov [Materials Development, Inc., Arlington Heights, IL 60004 (United States); Argonne National Laboratory, Argonne, IL 60439 (United States); Benmore, C.J. [Argonne National Laboratory, Argonne, IL 60439 (United States); Department of Physics, Arizona State University, AZ 85287 (United States); Tailor, A.N.; Tumber, S.K. [Materials Development, Inc., Arlington Heights, IL 60004 (United States); Neuefeind, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherry, B. [Magnetic Resonance Research Center, Arizona State University, AZ 85287 (United States); Department of Chemistry and Biochemistry, Arizona State University, AZ 85287 (United States); Yarger, J.L. [Department of Physics, Arizona State University, AZ 85287 (United States); Magnetic Resonance Research Center, Arizona State University, AZ 85287 (United States); Department of Chemistry and Biochemistry, Arizona State University, AZ 85287 (United States); Mou, Q. [Magnetic Resonance Research Center, Arizona State University, AZ 85287 (United States); Department of Chemistry and Biochemistry, Arizona State University, AZ 85287 (United States); Weber, W. [Department of Physics, Arizona State University, AZ 85287 (United States); Department of Chemistry and Biochemistry, Arizona State University, AZ 85287 (United States); Byrn, S.R. [Department of Industrial and Physical Pharmacy, Purdue University, IN 47907 (United States)

    2013-10-16

    Highlights: • Acoustic levitation was used to make phase-pure glassy forms of pharmaceutical compounds. • Neutrons, X-rays and NMR were used to characterize the glasses. • The glass comprised of slightly distorted molecules packed in a random network. • Potential for new drug synthesis routes is discussed. - Abstract: Acoustic levitation was used to trap 1–3 mm diameter drops of Probucol and other pharmaceutical materials in containerless conditions. Samples were studied in situ using X-ray diffraction and ex situ using neutron diffraction, NMR and DSC techniques. The materials were brought into non-equilibrium states by supersaturating solutions or by supercooling melts. The glass transition and crystallization temperatures of glassy Probucol were 29 ± 1 and 71 ± 1 °C respectively. The glassy form was stable with a shelf life of at least 8 months. A neutron/X-ray difference function of the glass showed that while molecular sub-groups remain rigid, many of the hydrogen correlations observed in the crystal become smeared out in the disordered material. The glass is principally comprised of slightly distorted Form I Probucol molecules with disordered packing rather than large changes in the individual molecular structure. Avoiding surface contact-induced nucleation provided access to highly non-equilibrium phases and enabled synthesis of phase-pure glasses.

  8. Profiling water content in soils with TDR: Comparison with the neutron probe technique

    International Nuclear Information System (INIS)

    Laurent, J.P.

    2000-01-01

    In November 1996, at a site on the Grenoble campus a 1.2-m-long neutron access tube, a 0.8-m fibreglass Trime access tube and three sets of 1-m twin-rod TDR probes were installed. Weekly measurements were made over a 9-month period. In addition, soil samples were taken from time to time with an auger, to determine gravimetric water-contents. The soil bulk density profile was initially characterised by gammametry using a Campbell TM probe. A Troxler TM 4300 was used for the neutron-probe measurements. The TDR signals, for further processing by TDR-SSI, were logged using a Trase 2000 from Soil Moisture Equipment Corporation TM . TDR methods were employed without any special calibration of the permittivity/water-content relationship: standard internal calibrations of the devices or Topp polynomial relation were always applied. The results of all these water-content profiling methods were compared in three ways: (i) the water-content profiles were plotted directly on the same graph for different dates; (ii) all the water contents measured at all dates and all depths were plotted against a corresponding 'reference', namely neutron probe or gravimetry; (iii) water balances were calculated for each method and their respective time-profiles analysed. There was fairly good agreement among the three profiling methods, indicating that TDR is now a viable alternative to nuclear techniques for soil water-content profiling. (author)

  9. Investigations on the comparator technique used in epithermal neutron activation analysis

    International Nuclear Information System (INIS)

    Bereznai, T.; Bodizs, D.; Keoemley, G.

    1977-01-01

    The possible extension of the comparator technique of reactor neutron activation analysis into the field of epithermal neutron activation has been investigated. Ruthenium was used for multi-isotopic comparator. Experiments show that conversion of the so-called reference k-factors - determined by irradiation with reactor neutrons - into ksup(epi)-factors usable at activation under cadmium filter, can be evaluated with fair accuracy. Sources and extent of errors and their contribution to the final error of analysis are discussed. For equal irradiation and counting times advantage of ENAA for several elements is obvious: the much lower background activity permitted the sample to be measured closer to the detector, under better geometry conditions, consequently, permitting several elements to be determined quantitatively. The number of elements determined and the sensitivity of the method are much dependent on the experimental conditions, especially on the composition of the sample, on the PHIsub(e) value, the irradiation time and the efficiency of the Ge(Li) detector. (T.G.)

  10. Deformation processes in functional materials studied by in situ neutron diffraction and ultrasonic techniques

    International Nuclear Information System (INIS)

    Sittner, P.; Novak, V.; Landa, M.; Lukas, P.

    2007-01-01

    The unique thermomechanical functions of shape memory alloys (hysteretic stress-strain-temperature responses) not their structural properties (as strength, fatigue, corrosion resistance, etc.) are primarily utilized in engineering applications. In order to better understand and predict the functional behavior, we have recently employed two dedicated non-invasive in situ experimental methods capable to follow the deformation/transformation processes in thermomechanically loaded polycrystalline samples. The in situ neutron diffraction method takes advantage of the ability of thermal neutrons to penetrate bulk samples. As a diffraction technique sensitive to interplanar spacings in crystalline solids, it provides in situ information on the changes in crystal structure, phase composition, phase stress and texture in the transforming samples. The combined in situ ultrasonic and electric resistance method follows variations of the electric resistance as well as speed and attenuation of acoustic waves propagating through the transforming sample. The acoustic waves are mainly sensitive to changes of elastic properties accompanying the deformation/transformation processes. The latter method thus follows the changes in interatomic bonds rather than changes in the interplanar lattice spacings focused in the neutron diffraction method. The methods are thus complementary. They are briefly described and selected experimental results obtained recently on NiTi alloys are presented and discussed

  11. Enhancements and Health-Related Studies of Neutron Activation Analysis Technique

    International Nuclear Information System (INIS)

    Soliman, M.A.M.

    2012-01-01

    The work presented in this thesis covers two major points. One algorithm concerns with establishment of an accurate standardization method with multi-elemental capabilities and low workload suitable for NAA standardization at ETRR-2. The second one deals with constructing and developing an effective nondestructive technique for analysis of liquid samples based on NAA using (very) short-lived radionuclides. To achieve the first goal, attention has been directed toward implementation of the k 0 -method for calculation of the elements concentrations in the samples. The k 0 -method of NAA standardization has a considerable success as a method for accurate multi-elemental analysis with comparable low workload. The k 0 - method is based on the fact that the unknown sample is irradiated with only one standard element as comparator. To access the implementation of this method at ETRR-2, careful and complete characterization of the neutron flux parameters in the irradiation positions as well as the efficiency calibration of the γ-ray spectrometer must be carried out. The required neutron flux parameters are: the ratio of the thermal to epithermal neutron fluxes (f) and the deviation factor (α) of the epithermal neutron flux from the ideal 1/E law. The work presented in Chapter 4 shows the efficiency calibration curve of the γ ray spectrometer system which was obtained using standard radioactive point sources. Moreover, the f and α parameters were determined in some selected irradiation sites using sets of Zr-Au as neutron flux monitors. Due to different locations relative to the reactor core, the available neutron fluxes in the selected irradiation positions differ substantially, so that different irradiation demands can be satisfied. The reference materials coal NIST 1632c and IAEA-Soil 7 were analyzed for data validation and good agreement between the experimental values and the certified values was obtained. The obtained results have revealed that the k 0 -NAA

  12. Combined neutron activation analysis techniques for multiple purposes at Portuguese research reactor

    International Nuclear Information System (INIS)

    Dung, H.M.; Freitas, M.C.; Beasley, D.; Almeida, S.M.; Dionisio, I; Canha, N.H.; Galinha, C.; Marques, J.G.

    2010-01-01

    Full text: Developments of the neutron activation analysis (NAA) techniques using Compton suppression system (CSS), fast pneumatic irradiation facility (SIPRA), epithermal neutron and automatic sample changers (ASCs) associated with the traditional NAA for trace element determination in various sample types are described with reference to specific conditions at the 1 MW Portuguese research reactor (RPI). Experiences in application of k o -IAEA software for data processing in order to deduce the results are also discussed. A selected number of sample types which are intended to the application in biological and environmental areas as well as industrial and material samples are demonstrated which provide challenges in the irradiation, measurement and the interpretation of data to which in most cases a combined solution should be made. The role that each NAA technique can play in the combined scheme along with their optimized characteristics has been studied and shown. The combined NAA techniques at RPI established for on-going and potential projects as well as analysis service with respect to the element scope (48), typically Ag, Al, As, Au, Ba, Br, Ca, Cd, Ce, CI, Co, Cr, Cs, Cu, Dy, Er, Eu, F, Fe, Ga, Hf, Hg, I, In, K, La, Mg, Mn, Mo, Na, Rb, Sb, Sc, Se, Si, Sm, Sn, Sr, Ta, Tb, Th, Ti, U, V, W, Vb, Zn and Zr along with detection limits, accuracies and precision's have been evaluated as a trace analysis method meeting the requirements of the intended applications

  13. Neutron time-of-flight techniques for investigation of the extinction effect

    International Nuclear Information System (INIS)

    Niimura, N.; Tomiyoshi, S.; Takahashi, J.; Harada, J.

    1975-01-01

    An application of the time-of-flight neutron diffraction technique to an investigation of the nature of the extinction effect in a single-crystal specimen is given. It is shown that the wavelength dependence of the extinction can be easily obtained by changing the scattering angle. An estimation of the extinction factor for a CuCl single crystal is given as an example and a comparison of the results with recent extinction theory [Becker and Coppens. Acta Cryst.(1974). A30, 129-147; 148-153] is made. (Auth.)

  14. Power measurement of the RA-3 reactor using the neutron noise technique and 16N

    International Nuclear Information System (INIS)

    Gomez, Angel

    2003-01-01

    This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16 N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

  15. Elemental investigation of talcum baby powder by X-Ray florescence and fast neutron activation Techniques

    International Nuclear Information System (INIS)

    Hassan, M. F.; Abd El Wahab, M.; Nada, A.

    2008-01-01

    Different samples of Egyptian and Hungarian talcum powders were studied, using X-ray florescence (XRF) and Fast Neutron Activation Analysis (FNAA) techniques to ensure the safety of its use. The K (X-rays) and the gamma-rays were measured, using Si(Li) and high-purity germanium (HPGe) spectrometers to detect and determine qualitatively and quantitatively the constituents of the studied samples. The concentrations of the elements (Mg, Si, Al, Fe, Zn, and Ba) were measured and their presence was confirmed by X-ray, lifetime and/or XRF measurements. One of these samples was also studied, using the Environmental Scanning Electron Microscope (ESEM)

  16. Investigation of Lecturer's Chalk by x-ray Florescence and Fast Neutron Activation Techniques

    International Nuclear Information System (INIS)

    Hassan, M.F.

    2011-01-01

    Different samples of lecturer's chalk were studied, using X-ray florescence (XRF) and Fast Neutron Activation Analysis (FNAA) techniques to ensure the safety of its use. The K (X-rays) and the gamma-rays were measured, using Si(Li) and high-purity germanium (HPGe) spectrometers to detect and determine qualitatively and quantitatively the constituents of the studied samples. The concentrations of the elements (Ca and small traces of Al, Fe, Mg and Si) were measured and their presence was confirmed by gamma-ray, lifetime and/or XRF measurements.

  17. 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.)

  18. Studies of Some Parameters Affecting The Efficiency and Accuracy of The Neutron Activation Analysis Technique

    International Nuclear Information System (INIS)

    Abdel-Haleem, A.S.; Zaghloul, R.A.; Sroor, A.; Abdel Sabour, M.F.

    2000-01-01

    The present studies deal with the optimum physical conditions which seriously affect the neutron activation analysis technique efficiency. An experimental work for the efficiency calibration of hyper pure germanium detectors especially for environmental studies is presented. This work showed that the tested parameters, under consideration, distance, mass and measured time, reveal a significant effect on the obtained data. These results, intern, affect the accuracy of the measurements. Further work on the test of other parameters is planned in our laboratory using special treatments and applying special computer programs

  19. Absolute absorbed dose measurements with an array of ionization chambers as part of a routine procedure of quality control for the VMAT technique

    International Nuclear Information System (INIS)

    Clemente Gutierrez, F.; Cabello Murillo, E.; Ramirez Ros, J. C.; Casa de Julian, M. A. de la

    2011-01-01

    Arcotheraphy techniques volumetric modulated (VMAT) treatments involve continuous variation of the gantry rotation speed, positions of the sheets and dose rate. Since all treatments are administered by continuous arcs, these techniques require quality control procedures to ensure quick and easy constancy of the calibration factor (total absorbed dose) for any gantry angle. We report here a simple method of quality control for the measurement of the calibration factor using an array of ionization chambers. The measurements were performed on a unit of 6 MV Elekta Synergy with VMAT, belonging to the Radiation Oncology service of the Defense Central Hospital Gomez Ulla.

  20. Estimation of organ-absorbed radiation doses during 64-detector CT coronary angiography using different acquisition techniques and heart rates: a phantom study

    Energy Technology Data Exchange (ETDEWEB)

    Matsubara, Kosuke; Koshida, Kichiro; Kawashima, Hiroko (Dept. of Quantum Medical Technology, Faculty of Health Sciences, Kanazawa Univ., Kanazawa (Japan)), email: matsuk@mhs.mp.kanazawa-u.ac.jp; Noto, Kimiya; Takata, Tadanori; Yamamoto, Tomoyuki (Dept. of Radiological Technology, Kanazawa Univ. Hospital, Kanazawa (Japan)); Shimono, Tetsunori (Dept. of Radiology, Hoshigaoka Koseinenkin Hospital, Hirakata (Japan)); Matsui, Osamu (Dept. of Radiology, Faculty of Medicine, Kanazawa Univ., Kanazawa (Japan))

    2011-07-15

    Background: Though appropriate image acquisition parameters allow an effective dose below 1 mSv for CT coronary angiography (CTCA) performed with the latest dual-source CT scanners, a single-source 64-detector CT procedure results in a significant radiation dose due to its technical limitations. Therefore, estimating the radiation doses absorbed by an organ during 64-detector CTCA is important. Purpose: To estimate the radiation doses absorbed by organs located in the chest region during 64-detector CTCA using different acquisition techniques and heart rates. Material and Methods: Absorbed doses for breast, heart, lung, red bone marrow, thymus, and skin were evaluated using an anthropomorphic phantom and radiophotoluminescence glass dosimeters (RPLDs). Electrocardiogram (ECG)-gated helical and ECG-triggered non-helical acquisitions were performed by applying a simulated heart rate of 60 beats per minute (bpm) and ECG-gated helical acquisitions using ECG modulation (ECGM) of the tube current were performed by applying simulated heart rates of 40, 60, and 90 bpm after placing RPLDs on the anatomic location of each organ. The absorbed dose for each organ was calculated by multiplying the calibrated mean dose values of RPLDs with the mass energy coefficient ratio. Results: For all acquisitions, the highest absorbed dose was observed for the heart. When the helical and non-helical acquisitions were performed by applying a simulated heart rate of 60 bpm, the absorbed doses for heart were 215.5, 202.2, and 66.8 mGy for helical, helical with ECGM, and non-helical acquisitions, respectively. When the helical acquisitions using ECGM were performed by applying simulated heart rates of 40, 60, and 90 bpm, the absorbed doses for heart were 178.6, 139.1, and 159.3 mGy, respectively. Conclusion: ECG-triggered non-helical acquisition is recommended to reduce the radiation dose. Also, controlling the patients' heart rate appropriately during ECG-gated helical acquisition with

  1. Absorbed dose by a CMOS in radiotherapy

    International Nuclear Information System (INIS)

    Borja H, C. G.; Valero L, C. Y.; Guzman G, K. A.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L. C.

    2011-10-01

    Absorbed dose by a complementary metal oxide semiconductor (CMOS) circuit as part of a pacemaker, has been estimated using Monte Carlo calculations. For a cancer patient who is a pacemaker carrier, scattered radiation could damage pacemaker CMOS circuits affecting patient's health. Absorbed dose in CMOS circuit due to scattered photons is too small and therefore is not the cause of failures in pacemakers, but neutron calculations shown an absorbed dose that could cause damage in CMOS due to neutron-hydrogen interactions. (Author)

  2. Synthesis of poly (acrylamide-co-metacrylic acid) hydrogels By means of gamma irradiation techniques: influence of Absorbed dose on the swelling process

    International Nuclear Information System (INIS)

    Rapado, Manuel; Altanes, Sonia; Sainz, Dianelys; Prado, Stalina

    1999-01-01

    In this report gamma radiation techniques were performed a double function of proceeding the processes of polymerization and crosslinking with the advantage of avoid the uses of chemicals crosslinks. The influence of absorbed dose on the swelling ratio as a function of pH have been presented. For these hydrogels, swelling studies indicated that swelling decrease with the increase of the absorbed dose from 10 to 50 kGy. It was confirmed that at the firsts stages (100-150 min) the diffusion studies were in accordance with Fickian behavior and the diffusion coefficients were obtained, whereas the latest stages were in good agreement with second-order diffusion kinetics proposed by Schott 1 .These news hydrogels exhibit a higher degree of swelling, a factor that, a priori, assures high biocompatibility because it increases the similarity with living tissues

  3. Quality assurance technique for absorbed dose distribution in external radiation therapy with non-physical wedges in consideration of the character of the imaging plate

    International Nuclear Information System (INIS)

    Fujibuchi, Toshioh; Fujisaki, Tatsuya; Kurokawa, Masayuki; Sakae, Takeji

    2011-01-01

    The film dose distribution method is used for profile measurements of non-physical wedges, because three-dimensional automatic control water phantoms cannot be used. Recently, many hospitals have adopted computed radiography (CR) systems in preference to automatic developing processors and films. This may allow use of automatic processors to be discontinued. In this study, a beam was irradiated to an Imaging Plate (IP), and then IP was exposed to a fixed amount of light with fading, and we then measured the off-center ratio (OCR) absorbed dose distribution in external radiation therapy with non-physical wedge. This was compared with the OCR measured with an ionization chamber dosimeter. It was consequently possible for IP to approximate the value measured by the ionization chamber dosimeter by using a metal filter. This method offers a simple quality assurance technique for absorbed dose distribution in external radiation therapy with non-physical wedges in consideration of the character of the IP. (author)

  4. TRU-ART: A cost-effective prototypical neutron imaging technique for transuranic waste certification systems

    International Nuclear Information System (INIS)

    Horton, W.S.

    1989-01-01

    The certification of defense radioactive waste as either transuranic or low-level waste requires very sensitive and accurate assay instrumentation to determine the specific radioactivity within an individual waste package. An assay instrument that employs a new technique (TRU-ART), which can identify the location of the radioactive material within a waste package, was designed, fabricated, and tested to potentially enhance the certification of problem defense waste drums. In addition, the assay instrumentation has potential application in radioactive waste reprocessing and neutron tomography. The assay instrumentation uses optimized electronic signal responses from an array of boral- and cadmium-shielded polyethylene-moderated 3 H detector packages. Normally, thermal neutrons that are detected by 3 H detectors have very poor spatial dependency that may be used to determine the location of the radioactive material. However, these shielded-detector packages of the TRU-ART system maintain the spatial dependency of the radioactive material in that the point of fast neutron thermalization is immediately adjacent to the 3 H detector. The TRU-ART was used to determine the location of radioactive material within three mock-up drums (empty, peat moss, and concrete) and four actual waste drums. The TRU-ART technique is very analogous to emission tomography. The mock-up drum and actual waste drum data, which were collected by the TRU-ART, were directly input into a algebraic reconstruction code to produce three-dimensional isoplots. Finally, a comprehensive fabrication cost estimate of the fielded drum assay system and the TRU-ART system was determined, and, subsequently, these estimates were used in a cost-benefit analysis to compare the economic advantage of the respective systems

  5. Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

    Science.gov (United States)

    Vagena, E.; Theodorou, K.; Stoulos, S.

    2018-04-01

    Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

  6. Measuring background by the DIN-1M spectrometer using the oscillating absorbing screen method

    International Nuclear Information System (INIS)

    Glazkov, Yu.Yu.; Liforov, V.G.; Novikov, A.G.; Parfenov, V.A.; Semenov, V.A.

    1982-01-01

    Technique for measuring background by a double pulse slow neutron spectrometer is described. To measure the background on oscillating absorbing screen (OAS) periodically overlapping primary neutron beam at the input of a mechanical interrupter was used. During the overlapping monochromatic neutrons conditioned the effect are removed out of the beam and general background conditions are not practically applied. Screen oscillation permits to realize the condition of simultaneous measurement of effect and background neutrons. The optimal period of oscillations amounts to approximately 3 min. Analysis of neutron spectra scattered with different materials and corresponding background curves measured by means of the OAS technique shows that the share of monochromatic neutrons passing through the screen constitutes less than 1% of elastic peak and relative decrease of the total background level doesn't exceed 1.5-2%

  7. Acceleration techniques for the direct use of CAD-based geometry in fusion neutronics analysis

    International Nuclear Information System (INIS)

    Wilson, Paul P.H.; Tautges, Timothy J.; Kraftcheck, Jason A.; Smith, Brandon M.; Henderson, Douglass L.

    2010-01-01

    The Direct Accelerated Geometry Monte Carlo (DAGMC) software library offers a unique approach to performing neutronics analysis on CAD-based geometries of fusion systems. By employing a number of acceleration techniques, the ray-tracing operations that are fundamental to Monte Carlo radiation transport are implemented efficiently for direct use on the CAD-based solid model, eliminating the need to translate to the native Monte Carlo input language. By forming hierarchical trees of oriented bounding boxes, one for each facet that results from a high-fidelity tessellation of the model, the ray-tracing performance is adequate to permit detailed analysis of large complex systems. In addition to the reduction in human effort and improvement in quality assurance that is found in the translation approaches, the DAGMC approach also permits the analysis of geometries with higher order surfaces that cannot be represented by many native Monte Carlo radiation transport tools. The paper describes the various acceleration techniques and demonstrates the resulting capability in a real fusion neutronics analysis.

  8. Proposal for Analysis of the Safeguarded Nuclear Materials 235U and 239Pu by Delayed Neutrons Technique

    International Nuclear Information System (INIS)

    El-Mongy, S.A.

    2000-01-01

    This paper introduces, describes and initiates a very sensitive and rapid non-destructive technique to be used for analysis of the safeguarded nuclear materials 235 U and 239 Pu. The technique is based on fission of the nuclear material by neutrons and then measuring the delayed neutrons produced from the neutron rich fission products. By this technique, fissile isotope content ( 235 U) can be determined in the presence of the other fissile (e.g. 239 Pu) or fertile isotopes (e.g. 238 U) in fresh and spent fuel. The time consumed for analysis of bulk materials by this technique is only 4 minutes. The method is also used for analysis of uranium in rock, sediment, soil, meteorites, lunar, biological, urine, archaeological, zircon sand and seawater samples. The method enables uranium in a sample to be measured without respect to its oxidation state, organic and inorganic elements

  9. Application of magnetomechanical hysteresis modeling to magnetic techniques for monitoring neutron embrittlement and biaxial stress

    International Nuclear Information System (INIS)

    Sablik, M.J.; Kwun, H.; Rollwitz, W.L.; Cadena, D.

    1992-01-01

    The objective is to investigate experimentally and theoretically the effects of neutron embrittlement and biaxial stress on magnetic properties in steels, using various magnetic measurement techniques. Interaction between experiment and modeling should suggest efficient magnetic measurement procedures for determining neutron embrittlement biaxial stress. This should ultimately assist in safety monitoring of nuclear power plants and of gas and oil pipelines. In the first six months of this first year study, magnetic measurements were made on steel surveillance specimens from the Indian Point 2 and D.C. Cook 2 reactors. The specimens previously had been characterized by Charpy tests after specified neutron fluences. Measurements now included: (1) hysteresis loop measurement of coercive force, permeability and remanence, (2) Barkhausen noise amplitude; and (3) higher order nonlinear harmonic analysis of a 1 Hz magnetic excitation. Very good correlation of magnetic parameters with fluence and embrittlement was found for specimens from the Indian Point 2 reactor. The D.C. Cook 2 specimens, however showed poor correlation. Possible contributing factors to this are: (1) metallurgical differences between D.C. Cook 2 and Indian Point 2 specimens; (2) statistical variations in embrittlement parameters for individual samples away from the stated men values; and (3) conversion of the D.C. Cook 2 reactor to a low leakage core configuration in the middle of the period of surveillance. Modeling using a magnetomechanical hysteresis model has begun. The modeling will first focus on why Barkhausen noise and nonlinear harmonic amplitudes appear to be better indicators of embrittlement than the hysteresis loop parameters

  10. Multielement comparison of instrumental neutron activation analysis techniques using reference materials

    International Nuclear Information System (INIS)

    Ratner, R.T.; Vernetson, W.G.

    1995-01-01

    Several instrumental neutron activation analysis techniques (parametric, comparative, and k o -standardization) are evaluated using three reference materials. Each technique is applied to National Institute of Standards and Technology standard reference materials, SRM 1577a (Bovine Liver) and SRM 2704 (Buffalo River Sediment), and the United States Geological Survey standard BHVO-1 (Hawaiian Basalt Rock). Identical (but not optimum) irradiation, decay, and counting schemes are employed with each technique to provide a basis for comparison and to determine sensitivities in a routine irradiation scheme. Fifty-one elements are used in this comparison; however, several elements are not detected in the reference materials due to rigid analytical conditions (e.g. insufficient length of irradiation or activity for radioisotope of interest decaying below the lower limit of detection before counting interval). Most elements are normally distributed around certified or consensus values with a standard deviation of 10%. For some elements, discrepancies are observed and discussed. The accuracy, precision, and sensitivity of each technique are discussed by comparing the analytical results to consensus values for the Hawaiian Basalt Rock to demonstrate the diversity of multielement applications. (author) 4 refs.; 2 tabs

  11. Spectra and absorbed dose by photo-neutrons in a solid water mannequin exposed to a Linac of 15 MV; Espectros y dosis absorbida por fotoneutrones en un maniqui de agua solida expuesta a una Linac de 15 MV

    Energy Technology Data Exchange (ETDEWEB)

    Benites R, J. [Centro Estatal de Cancerologia de Nayarit, Servicio de Seguridad Radiologica, Calz. de la Cruz 118 Sur, 63000 Tepic, Nayarit (Mexico); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Apdo. Postal 336, 98000 Zacatecas (Mexico); Velazquez F, J., E-mail: jlbenitesr@prodigy.net.mx [Universidad Autonoma de Nayarit, Posgrado en Ciencias Biologico Agropecuarias, Carretera Tepic-Compostela Km 9, 63780 Jalisco-Nayarit (Mexico)

    2012-10-15

    Using Monte Carlo methods was modeled a solid water mannequin; according to the ICRU 44 (1989), Tissue substitutes in radiation dosimetry and measurements, of the International Commission on Radiation Units and Measurements; Report 44. This material Wt 1 is made of H (8.1%), C (67.2%), N (2.4%), O (19.9%), Cl (0.1%), Ca (2.3%) and its density is of 1.02 gr/cm{sup 3}. The mannequin was put instead of the patient, inside the treatment room and the spectra and absorbed dose were determined by photo-neutrons exposed to a Linac of 15 MV. (Author)

  12. Application of the coincidence counting technique to DD neutron spectrometry data at the NIF, OMEGA, and Z

    Energy Technology Data Exchange (ETDEWEB)

    Lahmann, B., E-mail: lahmann@mit.edu; Milanese, L. M.; Han, W.; Gatu Johnson, M.; Séguin, F. H.; Frenje, J. A.; Petrasso, R. D. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hahn, K. D.; Jones, B. [Sandia National Laboratory, Albuquerque, New Mexico 87123 (United States)

    2016-11-15

    A compact neutron spectrometer, based on a CH foil for the production of recoil protons and CR-39 detection, is being developed for the measurements of the DD-neutron spectrum at the NIF, OMEGA, and Z facilities. As a CR-39 detector will be used in the spectrometer, the principal sources of background are neutron-induced tracks and intrinsic tracks (defects in the CR-39). To reject the background to the required level for measurements of the down-scattered and primary DD-neutron components in the spectrum, the Coincidence Counting Technique (CCT) must be applied to the data. Using a piece of CR-39 exposed to 2.5-MeV protons at the MIT HEDP accelerator facility and DD-neutrons at Z, a significant improvement of a DD-neutron signal-to-background level has been demonstrated for the first time using the CCT. These results are in excellent agreement with previous work applied to DT neutrons.

  13. Analysis and development of deterministic and stochastic neutron noise computing techniques with applications to thermal and fast reactors

    International Nuclear Information System (INIS)

    Rouchon, Amelie

    2016-01-01

    Neutron noise analysis addresses the description of small time-dependent flux fluctuations induced by small global or local perturbations of the macroscopic cross-sections. These fluctuations may occur in nuclear reactors due to density fluctuations of the coolant, to vibrations of fuel elements, control rods, or any other structures in the core. In power reactors, ex-core and in-core detectors can be used to monitor neutron noise with the aim of detecting possible anomalies and taking the necessary measures for continuous safe power production. The objective of this thesis is to develop techniques for neutron noise analysis and especially to implement a neutron noise solver in the deterministic transport code APOLLO3 developed at CEA. A new Monte Carlo algorithm that solves the transport equations for the neutron noise has been also developed. In addition, a new vibration model has been developed. Moreover, a method based on the determination of a new steady state has been proposed for the linear and the nonlinear full theory so as to improve the traditional neutron noise theory. In order to test these new developments we have performed neutron noise simulations in one-dimensional systems and in a large pressurized water reactor with heavy baffle in two and three dimensions with APOLLO3 in diffusion and transport theories. (author) [fr

  14. Application of the coincidence counting technique to DD neutron spectrometry data at the NIF, OMEGA, and Z

    International Nuclear Information System (INIS)

    Lahmann, B.; Milanese, L. M.; Han, W.; Gatu Johnson, M.; Séguin, F. H.; Frenje, J. A.; Petrasso, R. D.; Hahn, K. D.; Jones, B.

    2016-01-01

    A compact neutron spectrometer, based on a CH foil for the production of recoil protons and CR-39 detection, is being developed for the measurements of the DD-neutron spectrum at the NIF, OMEGA, and Z facilities. As a CR-39 detector will be used in the spectrometer, the principal sources of background are neutron-induced tracks and intrinsic tracks (defects in the CR-39). To reject the background to the required level for measurements of the down-scattered and primary DD-neutron components in the spectrum, the Coincidence Counting Technique (CCT) must be applied to the data. Using a piece of CR-39 exposed to 2.5-MeV protons at the MIT HEDP accelerator facility and DD-neutrons at Z, a significant improvement of a DD-neutron signal-to-background level has been demonstrated for the first time using the CCT. These results are in excellent agreement with previous work applied to DT neutrons.

  15. Application of the coincidence counting technique to DD neutron spectrometry data at the NIF, OMEGA, and Z.

    Science.gov (United States)

    Lahmann, B; Milanese, L M; Han, W; Gatu Johnson, M; Séguin, F H; Frenje, J A; Petrasso, R D; Hahn, K D; Jones, B

    2016-11-01

    A compact neutron spectrometer, based on a CH foil for the production of recoil protons and CR-39 detection, is being developed for the measurements of the DD-neutron spectrum at the NIF, OMEGA, and Z facilities. As a CR-39 detector will be used in the spectrometer, the principal sources of background are neutron-induced tracks and intrinsic tracks (defects in the CR-39). To reject the background to the required level for measurements of the down-scattered and primary DD-neutron components in the spectrum, the Coincidence Counting Technique (CCT) must be applied to the data. Using a piece of CR-39 exposed to 2.5-MeV protons at the MIT HEDP accelerator facility and DD-neutrons at Z, a significant improvement of a DD-neutron signal-to-background level has been demonstrated for the first time using the CCT. These results are in excellent agreement with previous work applied to DT neutrons.

  16. Mapping sediment deposite on tank FB-901 using neutron back scattering technique

    International Nuclear Information System (INIS)

    Wibisono; Sugiharto; Zulkifli Lubis; Phyu Phyu Aung Myint; Thin Moe Hlaing

    2016-01-01

    Tank FB-901 is storage tank for temporary material production with a diameter 11 m and a high 12 m. This tank has been use about 10 years so it is suspected there is sediment in it. Neutron back scattering technique has been used to detected the level of sediment inside so it can be seen the volume of liquid properly and avoid problem in the nozzle outlet. AmBe neutron source with activity one Curie shoot into the tank to enable back scattering intensity from material. Measurement using He-3 detector, radiation counter Ludlum model 2200 scaler ratemeter and mechanical motor controlled by computer. Investigation were taken at around the tank from the bottom to the top on each step 50 mm height 8000 mm. Scan determined the distance between 500 mm and measurement time 3 seconds to each sample point. Investigation found the sediment level average 1000 mm by 1500 mm highest and lowest level 100 mm. Fluctuating liquid level observed maximum of 7800 mm and average of 7000 mm. Cleaning tank advised to avoid blockage of the nozzle and material volume is measured accurately. (author)

  17. Use of higher order signal moments and high speed digital sampling technique for neutron flux measurements

    Science.gov (United States)

    Baers, L. B.; Gutierrez, T. Rivero; Mendoza, R. A. Carrillo; Santana, G. Jimenez

    1993-08-01

    The second (conventional variance or Campbell signal), the third, and the modified fourth order central signal moments associated with the amplified and filtered currents from two electrodes of an ex-core neutron sensitive fission detector were measured versus the reactor power of the 1-MW TRIGA reactor in Mexico City. Two channels of a high-speed (400-MHz) multiplexing data sampler and an analog-to-digital converter with 12-b resolution and 1-Mword buffer memory were used. The data were further retrieved into a PC, and estimates for autocorrelation and cross-correlation moments up to the fifth order, coherence, skewness, excess, etc., quantities were calculated offline. Five-mode operation of the detector was achieved, including conventional counting rates and currents in agreement with theory and the authors' previous results with analog techniques. The signals are proportional to the neutron flux and reactor power in some flux ranges. The suppression of background noise is improved and the lower limit of the measurement range is extended as the order of moment is increased, in agreement with theory.

  18. 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.)

  19. Evaluation of the natural radioactivity using thermoluminescence, gamma spectrometry and neutron activation techniques

    International Nuclear Information System (INIS)

    Fausto, A.M.; Otsubo, S.M.; Paes, W.S.; Yoshimura, E.M.; Okuno, E.; Hiodo, F.; Marques, L.S.; Alcala, A.L.

    1996-01-01

    The evaluation of the radiation dose in the living species due to the natural radioactivity is the main objective of this paper. The region that had been monitored was Intrusive Suite of Itu near to Sao Paulo city. Lateral[and depth distributions of natural radioactivity of the soil were determined using the techniques of thermoluminescence gamma spectrometry and neutron[activation. From the concentration in the soil of K-40 and radioactive elements of U and Th series experimentally determined, the dose due to gamma rays in air at 1 m from the soil was evaluated and the results compared with the direct dose measured with the portable rate-meter. (authors). 11 refs., 3 figs

  20. Mechanisms of cement leaching and degradation - integration of neutron imaging techniques

    International Nuclear Information System (INIS)

    Payne, Timothy E.; Aldridge, Laurence P.; Brew, Daniel R.M.; McGlinn, Peter J.; De Beer, Frikkie C.; Radebe, Mabuti J.; Nshimirimana, Robert

    2012-01-01

    Cementitious material is a commonly used wasteform for low and intermediate level radioactive waste, and comprises a major part of both structural components and barriers in many repository concepts. When exposed to water, cement-based barriers and waste-forms are expected to degrade by mechanisms involving both chemical and structural changes. The research program addresses several aspects of these processes, including the leaching of the waste-forms, water transport properties, as well as the effect of high pH cement leachates on the chemical and physical properties of surrounding materials (including clay barriers and host regolith materials). Chemical leaching tests and analyses by techniques such as electron microscopy can be augmented by neutron radiography and tomography. These methods provide a useful non-destructive method of determining properties related to water transport in cementitious materials, in particular the sorptivity and pore size distribution

  1. Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques

    International Nuclear Information System (INIS)

    Jacqmin, R.P.; Finck, P.J.; Palmiotti, G.

    1994-01-01

    This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution

  2. Neutron activation analysis - an old technique but is it still useful?

    International Nuclear Information System (INIS)

    McOrist, G.D.

    1998-01-01

    Neutron activation analysis (NAA) is a non-destructive technique of analysis making it particularly suitable to analyse certain types of samples or particular elements, such as diamonds, black precious opal and ancient fish fossil bones. It is particularly suited to analysing solid samples since no sample dissolution, with its associated problems, prior to analysis is necessary. NAA is remarkably free of matrix effects essentially eliminating complex pre analysis procedures. It is uniquely suited to determining low levels of certain elements such as Cl, Br and I, particularly in organic samples. Its ability to determine sub-ppm levels of Cl in transformer oils has greatly assisted in the development of a patented method for the destruction of PCB ' s. It is possible to analyse large samples in order to overcome homogeneity problems, which is particularly useful in the precious metals industry. Radiochemical NAA enables the determination of very low levels of particular elements (eg Ta) in samples traditionally difficult to dissolve (author)

  3. Development of Large Sample Neutron Activation Technique for New Applications in Thailand

    International Nuclear Information System (INIS)

    Laoharojanaphand, S.; Tippayakul, C.; Wonglee, S.; Channuie, J.

    2018-01-01

    The development of the Large Sample Neutron Activation Analysis (LSNAA) in Thailand is presented in this paper. The technique had been firstly developed with rice sample as the test subject. The Thai Research Reactor-1/Modification 1 (TRR-1/M1) was used as the neutron source. The first step was to select and characterize an appropriate irradiation facility for the research. An out-core irradiation facility (A4 position) was first attempted. The results performed with the A4 facility were then used as guides for the subsequent experiments with the thermal column facility. The characterization of the thermal column was performed with Cu-wire to determine spatial distribution without and with rice sample. The flux depression without rice sample was observed to be less than 30% while the flux depression with rice sample increased to within 60%. The flux monitors internal to the rice sample were used to determine average flux over the rice sample. The gamma selfshielding effect during gamma measurement was corrected using the Monte Carlo simulation. The ratio between the efficiencies of the volume source and the point source for each energy point was calculated by the MCNPX code. The research team adopted the k0-NAA methodology to calculate the element concentration in the research. The k0-NAA program which developed by IAEA was set up to simulate the conditions of the irradiation and measurement facilities used in this research. The element concentrations in the bulk rice sample were then calculated taking into account the flux depression and gamma efficiency corrections. At the moment, the results still show large discrepancies with the reference values. However, more research on the validation will be performed to identify sources of errors. Moreover, this LS-NAA technique was introduced for the activation analysis of the IAEA archaeological mock-up. The results are provided in this report. (author)

  4. Review of personal monitoring techniques for the measurement of absorbed dose from external beta and low energy photon radiation

    DEFF Research Database (Denmark)

    Christensen, Poul

    1986-01-01

    The techniques available at present for personal monitoring of doses from external beta and low energy photon radiation are reviewed. The performance of currently used dosimetry systems is compared with that recommended internationally, and developments for improving the actual performance...

  5. Reducing the absorbed dose in analogue radiography of infant chest images by improving the image quality, using image processing techniques

    International Nuclear Information System (INIS)

    Karimian, A.; Yazdani, S.; Askari, M. A.

    2011-01-01

    Radiographic inspection is one of the most widely employed techniques for medical testing methods. Because of poor contrast and high un-sharpness of radiographic image quality in films, converting radiographs to a digital format and using further digital image processing is the best method of enhancing the image quality and assisting the interpreter in their evaluation. In this research work, radiographic films of 70 infant chest images with different sizes of defects were selected. To digitise the chest images and employ image processing the two algorithms (i) spatial domain and (ii) frequency domain techniques were used. The MATLAB environment was selected for processing in the digital format. Our results showed that by using these two techniques, the defects with small dimensions are detectable. Therefore, these suggested techniques may help medical specialists to diagnose the defects in the primary stages and help to prevent more repeat X-ray examination of paediatric patients. (authors)

  6. Neutron beam applications - Development of single crystal structure analysis technique using the HANARO neutron four circle diffractometer

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Il Hwan; Kim, Moon Jib; Kim, Jin Gyu [Chungnam National University, Taejon (Korea)

    2000-04-01

    As the four circle diffractometer (FCD) has been set up in HANARO, it has become possible to study the single crystal structures by means of the neutron diffraction. Taking account of the geometry of the FCD, a program for the control of te FCD and neutron data acquisition operating under Windows' circumstance has been accomplished. Also, a computer program which can automatically measure the diffraction intensity data has been developed. All data obtained from the FCD are processed automatically for further work and a software for the single crystal structure analyses has been prepared. A KC1 single crystal was selected as first test sample for a structure analysis had been successfully performed on the FCD using in-house developed program and accordingly their functionings with precision were confirmed. For regular single crystal diffraction experiments, the structure analyses of chrysoberyl and Zr(Y)0{sub 1.87} single crystals were performed using both neutron and X-ray diffraction methods, and the result showed that the neutron diffraction work is superior to the X-ray one from the viewpoint of certain crystallographic information obtainable only from the former one. 24 refs., 15 figs., 15 tabs. (Author)

  7. 4π-spectrometer technique for measurements of secondary neutron average number in nuclear fission by 252Cf neutrons

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.A.; Barashkov, Yu.A.; Golovanov, O.A.; Sidorov, L.V.

    1977-01-01

    A method for determining the average number of secondary neutrons anti ν produced in nuclear fission by the neutrons of the 252 Cf fission spectra by means of a 4π time-of-flight spectrometer is described. Layers of 252 Cf and an isotope studied are placed close to each other; if the isotope layer density is 1 mg/cm 2 probability of its fission is about 10 -5 per one spontaneous fission of californium. Fission fragments of 252 Cf and the isotope investigated have been detected by two surface-barrier counters with an efficiency close to 100%. The layers and the counters are situated in a measuring chamber placed in the center of the 4π time-of-flight spectrometer. The latter is utilized as a neutron counter because of its fast response. The method has been verified by carrying out measurements for 235 U and 239 Pu. A comparison of the experimental and calculated results shows that the method suggested can apply to determine the number of secondary neutrons in fission of isotopes that have not been investigated yet

  8. Validation of computational methods for treatment planning of fast-neutron therapy using activation foil techniques

    International Nuclear Information System (INIS)

    Nigg, D.W.; Wemple, C.A.; Hartwell, J.K.; Harker, Y.D.; Venhuizen, J.R.; Risler, R.

    1997-12-01

    A closed-form direct method for unfolding neutron spectra from foil activation data is presented. The method is applied to measurements of the free-field neutron spectrum produced by the proton-cyclotron-based fast-neutron radiotherapy facility at the University of Washington (UW) School of Medicine. The results compare favorably with theoretical expectations based on an a-priori calculational model of the target and neutron beamline configuration of the UW facility

  9. Neutronics experiments, radiation detectors and nuclear techniques development in the EU in support of the TBM design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M., E-mail: maurizio.angelone@enea.it [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Fischer, U. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Flammini, D. [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Jodlowski, P. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Klix, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kodeli, I. [Jožef Stefan Institute, Ljubljana (Slovenia); Kuc, T. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Leichtle, D. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lilley, S. [Culham Centre for Fusion Energy, Culham, OX14 3DB (United Kingdom); Majerle, M.; Novák, J. [Nuclear Physics Institute of the ASCR, Řež 130, 250 68 Řež (Czech Republic); Ostachowicz, B. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Packer, L.W. [Culham Centre for Fusion Energy, Culham, OX14 3DB (United Kingdom); Pillon, M. [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Pohorecki, W. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Radulović, V. [Jožef Stefan Institute, Ljubljana (Slovenia); Šimečková, E. [Nuclear Physics Institute of the ASCR, Řež 130, 250 68 Řež (Czech Republic); and others

    2015-10-15

    Highlights: • A number of experiments and tests are ongoing to develop detectors and methods for HCLL and HCPM ITER-TBM. • Experiments for measuring gas production relevant to IFMIF are also performed using a cyclotron. • A benchmark experiment with a Cu block is performed to validate copper cross sections. • Experimental techniques to measure tritium in TBM are presented. • Experimental verification of activation cross sections for a Neutron Activation System for TBM is addressed. - Abstract: The development of high quality nuclear data, radiation detectors and instrumentation techniques for fusion technology applications in Europe is supported by Fusion for Energy (F4E) and conducted in a joint and collaborative effort by several European research associations (ENEA, KIT, JSI, NPI, AGH, and CCFE) joined to form the “Consortium on Nuclear Data Studies/Experiments in Support of TBM Activities”. This paper presents the neutronics activities carried out by the Consortium. A selection of available results are presented. Among then a benchmark experiment on a pure copper block to study the Cu cross sections at neutron energies relevant to fusion, the fabrication of prototype neutron detectors able to withstand harsh environment and temperature >200 °C (artificial diamond and self-powered detectors) developed for operating in ITER-TBM as well as measurement of relevant activation and integral gas production cross-sections. The latter measured at neutron energies relevant to IFMIF (>14 MeV) and the development of innovative experimental techniques for tritium measurement in TBM.

  10. Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

    Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

  11. Implementation of neutron diffraction technique at Nuclear Center of National Institute of Nuclear Research for study of materials

    International Nuclear Information System (INIS)

    Macias Betanzos, L.R.

    1993-01-01

    The Neutron Diffraction technique, it's a helpful tool for the study of materials. The purpose, was to verify that such technique works with the Neutron Diffractometer of National Institute of Nuclear Research. The scope, is to study crystalline materials by the Neutron Diffraction Method, that means it completion with Bragg's Law. There exist a lot of diffraction techniques that depend on the kind of study to do. In this case the study was to measure known samples to have a correlation between parameters such a extinction factor and dislocation density. Known copper deformed samples were measured to observe the extinction effect and it could be observed. We had to calibrate the Neutron Diffractometer, the detection system and to have an optimal movement control of diffractometer devices by mean of a microcomputer. Also, was necessary to control the Reactor TRIGA operation to minimize the neutron flux oscillation. It was not possible the quantification of dislocation density in the samples because the relation signal/background was about one and it gives high inaccuracy. To correct this problem, it's necessary to have a better shielding to minimize the contribution of the background. The conclusion is that the Neutron Diffractometer is in conditions to carry out investigation on the material field, today it can be lattice constants, crystalline phases and measurements of metallic textures. For such studies, it's necessary to have samples with 2 cm 3 or higher to increase the relation signal/background. At present, we have the process software to give the interpretation of the Neutron Diffraction process. (Author). 12 refs, 16 figs

  12. A comparison of analogue and digital techniques in upper gastrointestinal examinations: absorbed dose and diagnostic quality of the images

    Energy Technology Data Exchange (ETDEWEB)

    Axelsson, B.; Hansson, I.B. [Department of Hospital Physics, Karolinska Hospital, Stockholm (Sweden); Boden, K.; Witt, H.H. [Department of Diagnostic Radiology, Karolinska Hospital, Stockholm (Sweden); Fransson, S.G. [Department of Thoracic Radiology, University Hospital, Linkoeping (Sweden); Persliden, J. [Department of Radiation Physics, University Hospital, Linkoeping (Sweden)

    2000-08-01

    This study was performed to investigate whether patient exposure and diagnostic quality of the image is significantly influenced by the introduction of digital image acquisition techniques. Evaluation was performed for three different techniques (analogue, analogue fluoro+digital radiography, digital) in examination of the upper gastrointestinal tract. The evaluation was done from data acquired in three different departments. Patient exposure was recorded as KERMA-area product (KAP) and the individual patient readings were normalised to a standard size patient. Image quality was assessed using visual grading with a reference image. The recorded KAP values were significantly higher (22.3 Gycm{sup 2}) for the fully digital technique compared to the others (analogue 6.8 Gycm{sup 2}, analogue+digital 3.6 Gycm{sup 2}). This was due mostly to an increased number of exposures. The diagnostic quality of the image was, however, also regarded to be slightly lower for the technique giving the lowest patient dose with the smallest number of exposures (analogue+digital). The digital examination technique, as used in this study, thus resulted in significantly higher patient dose without any significant gain in diagnostic quality of the image. (orig.)

  13. A comparison of analogue and digital techniques in upper gastrointestinal examinations: absorbed dose and diagnostic quality of the images

    International Nuclear Information System (INIS)

    Axelsson, B.; Hansson, I.B.; Boden, K.; Witt, H.H.; Fransson, S.G.; Persliden, J.

    2000-01-01

    This study was performed to investigate whether patient exposure and diagnostic quality of the image is significantly influenced by the introduction of digital image acquisition techniques. Evaluation was performed for three different techniques (analogue, analogue fluoro+digital radiography, digital) in examination of the upper gastrointestinal tract. The evaluation was done from data acquired in three different departments. Patient exposure was recorded as KERMA-area product (KAP) and the individual patient readings were normalised to a standard size patient. Image quality was assessed using visual grading with a reference image. The recorded KAP values were significantly higher (22.3 Gycm 2 ) for the fully digital technique compared to the others (analogue 6.8 Gycm 2 , analogue+digital 3.6 Gycm 2 ). This was due mostly to an increased number of exposures. The diagnostic quality of the image was, however, also regarded to be slightly lower for the technique giving the lowest patient dose with the smallest number of exposures (analogue+digital). The digital examination technique, as used in this study, thus resulted in significantly higher patient dose without any significant gain in diagnostic quality of the image. (orig.)

  14. Sound Absorbers

    Science.gov (United States)

    Fuchs, H. V.; Möser, M.

    Sound absorption indicates the transformation of sound energy into heat. It is, for instance, employed to design the acoustics in rooms. The noise emitted by machinery and plants shall be reduced before arriving at a workplace; auditoria such as lecture rooms or concert halls require a certain reverberation time. Such design goals are realised by installing absorbing components at the walls with well-defined absorption characteristics, which are adjusted for corresponding demands. Sound absorbers also play an important role in acoustic capsules, ducts and screens to avoid sound immission from noise intensive environments into the neighbourhood.

  15. Neutron physical investigations on the shutdown effect of small boronated absorbing spheres for pebble-bed high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Sgouridis, S.; Schurrer, F.; Muller, H.; Ninaus, W.; Oswald, K.; Neef, R.D.; Schaal, H.

    1987-01-01

    An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres

  16. Non destructive multi elemental analysis using prompt gamma neutron activation analysis techniques: Preliminary results for concrete sample

    Energy Technology Data Exchange (ETDEWEB)

    Dahing, Lahasen Normanshah [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency (Nuklear Malaysia), Bangi 43000, Kajang (Malaysia); Yahya, Redzuan [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Yahya, Roslan; Hassan, Hearie [Malaysian Nuclear Agency (Nuklear Malaysia), Bangi 43000, Kajang (Malaysia)

    2014-09-03

    In this study, principle of prompt gamma neutron activation analysis has been used as a technique to determine the elements in the sample. The system consists of collimated isotopic neutron source, Cf-252 with HPGe detector and Multichannel Analysis (MCA). Concrete with size of 10×10×10 cm{sup 3} and 15×15×15 cm{sup 3} were analysed as sample. When neutrons enter and interact with elements in the concrete, the neutron capture reaction will occur and produce characteristic prompt gamma ray of the elements. The preliminary result of this study demonstrate the major element in the concrete was determined such as Si, Mg, Ca, Al, Fe and H as well as others element, such as Cl by analysis the gamma ray lines respectively. The results obtained were compared with NAA and XRF techniques as a part of reference and validation. The potential and the capability of neutron induced prompt gamma as tool for multi elemental analysis qualitatively to identify the elements present in the concrete sample discussed.

  17. Trace elements in California aerosols. Part I. Instrumental neutron activation analysis techniques

    International Nuclear Information System (INIS)

    Ragaini, R.C.; Ralston, H.R.; Garvis, D.; Kaifer, R.

    1975-01-01

    Instrumental Neutron Activation Analysis (INAA) done at LLL played a key role in the 1972--1974 California Aerosol Characterization Experiment (ACHEX), a major experiment in the chemistry of aerosols in urban and non-urban sites of California sponsored by the State of California Air Resources Board. The main purpose of INAA was to measure the particle size distributions and diurnal patterns of key chemical constituents in aerosols collected in California. These data were used to satisfy some of the key objectives of ACHEX, including aerosol characterization and evaluation of the origins and evolutions of aerosols. Secondary uses of INAA were the validations of the Lundgren rotating drum cascade impactors used in the ACHEX, and validations of other analytical techniques used in the chemical analyses. As a result of these studies, it was concluded that techniques using INAA were useful operational methods for chemical analysis of aerosols collected over two-hour periods in urban air with an active monitoring program. (U.S.)

  18. Study of water permeability in concrete by neutron and gamma-ray techniques

    International Nuclear Information System (INIS)

    Abd El-Monem, A.M.M.

    2010-01-01

    water infiltration in various building materials , namely concrete used for buildings basement and underwater construction is the main concern of the studies performed in this thesis. The studies aim to develop a nuclear techniques for investigation a concrete mixes with different additives capable to decrease concrete porosity and intern resist water propagation inside concrete materials without any deterioration of concrete physical and mechanical properties . These issues were achieved through the preparation of ordinary concrete mixes with different percentages of silica fume. Concrete samples of different shape and geometries were made to study water diffusion when the concrete samples are submerged in water for different periods of time. The concrete samples were first sealed by molten asphalt from all sides expect two opposite faces to ensure water migration only along one direction. Water infiltration in concrete samples with different percentages of silica fume and submerged in tap and seawater for different periods of time was studied by neutrons and gamma techniques. Also, water propagation in mortar samples with different percentages of silica fume was studied by electrical methods based on measuring the variation in electrical conductivity of these samples.

  19. A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk

    1994-08-01

    The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author)

  20. Pulsed-laser time-resolved thermal mirror technique in low-absorbance homogeneous linear elastic materials.

    Science.gov (United States)

    Lukasievicz, Gustavo V B; Astrath, Nelson G C; Malacarne, Luis C; Herculano, Leandro S; Zanuto, Vitor S; Baesso, Mauro L; Bialkowski, Stephen E

    2013-10-01

    A theoretical model for a time-resolved photothermal mirror technique using pulsed-laser excitation was developed for low absorption samples. Analytical solutions to the temperature and thermoelastic deformation equations are found for three characteristic pulse profiles and are compared to finite element analysis methods results for finite samples. An analytical expression for the intensity of the center of a continuous probe laser at the detector plane is derived using the Fresnel diffraction theory, which allows modeling of experimental results. Experiments are performed in optical glasses, and the models are fitted to the data. The parameters of the fit are in good agreement with previous literature data for absorption, thermal diffusion, and thermal expansion of the materials tested. The combined modeling and experimental techniques are shown to be useful for quantitative determination of the physical properties of low absorption homogeneous linear elastic material samples.

  1. Experimental validation of energy parameters in parabolic trough collector with plain absorber and analysis of heat transfer enhancement techniques

    Science.gov (United States)

    Bilal, F. R.; Arunachala, U. C.; Sandeep, H. M.

    2018-01-01

    The quantum of heat loss from the receiver of the Parabolic Trough Collector is considerable which results in lower thermal efficiency of the system. Hence heat transfer augmentation is essential which can be attained by various techniques. An analytical model to evaluate the system with bare receiver performance was developed using MATLAB. The experimental validation of the model resulted in less than 5.5% error in exit temperature using both water and thermic oil as heat transfer fluid. Further, heat transfer enhancement techniques were incorporated in the model which included the use of twisted tape inserts, nanofluid, and a combination of both for further enhancement. It was observed that the use of evacuated glass cover in the existing setup would increase the useful heat gain up to 5.3%. Fe3O4/H2O nanofluid showed a maximum enhancement of 56% in the Nusselt number for the volume concentration of 0.6% at highest Reynolds number. Similarly, twisted tape turbulators (with twist ratio of 2) taken alone with water exhibited 59% improvement in Nusselt number. Combining both the heat transfer augmentation techniques at their best values revealed the Nusselt number enhancement up to 87%. It is concluded that, use of twisted tape with water is the best method for heat transfer augmentation since it gives the maximum effective thermal efficiency amongst all for the range of Re considered. The first section in your paper

  2. New neutron imaging techniques to close the gap to scattering applications

    International Nuclear Information System (INIS)

    Lehmann, Eberhard H.; Peetermans, S.; Trtik, P.; Betz, B.; Grünzweig, C.

    2017-01-01

    Neutron scattering and neutron imaging are activities at the strong neutron sources which have been developed rather independently. However, there are similarities and overlaps in the research topics to which both methods can contribute and thus useful synergies can be found. In particular, the spatial resolution of neutron imaging has improved recently, which - together with the enhancement of the efficiency in data acquisition- can be exploited to narrow the energy band and to implement more sophisticated methods like neutron grating interferometry. This paper provides a report about the current options in neutron imaging and describes how the gap to neutron scattering data can be closed in the future, e.g. by diffractive imaging, the use of polarized neutrons and the dark-field imagining of relevant materials. This overview is focused onto the interaction between neutron imaging and neutron scattering with the aim of synergy. It reflects mainly the authors’ experiences at their PSI facilities without ignoring the activities at the different other labs world-wide. (paper)

  3. New neutron imaging techniques to close the gap to scattering applications

    Science.gov (United States)

    Lehmann, Eberhard H.; Peetermans, S.; Trtik, P.; Betz, B.; Grünzweig, C.

    2017-01-01

    Neutron scattering and neutron imaging are activities at the strong neutron sources which have been developed rather independently. However, there are similarities and overlaps in the research topics to which both methods can contribute and thus useful synergies can be found. In particular, the spatial resolution of neutron imaging has improved recently, which - together with the enhancement of the efficiency in data acquisition- can be exploited to narrow the energy band and to implement more sophisticated methods like neutron grating interferometry. This paper provides a report about the current options in neutron imaging and describes how the gap to neutron scattering data can be closed in the future, e.g. by diffractive imaging, the use of polarized neutrons and the dark-field imagining of relevant materials. This overview is focused onto the interaction between neutron imaging and neutron scattering with the aim of synergy. It reflects mainly the authors’ experiences at their PSI facilities without ignoring the activities at the different other labs world-wide.

  4. Trace elements in high purity materials for advanced technology: contribution of neutron activation analysis and radioanalytical technique

    International Nuclear Information System (INIS)

    Gallorini, M.; Pietra, R.; Sabbioni, E.

    1991-01-01

    Neutron activation analysis and radioanalytical techniques have been employed to investigate problems related to trace elements and high purity technology materials. Applications of these techniques are overviewed: semiconductor technology as in the case of As and In ion implantation in high purity silicon; problems related to trace elements impurities in thermometric measurements; coating materials to prevent trace elements contamination in biological sampling and metals release from human prostheses. (author) 8 refs.; 2 figs.; 8 tabs

  5. Gamma spectroscopic studies of the neutron-deficient g-g nucleus 74Kr by means of a neutron multiplicity measurement technique

    International Nuclear Information System (INIS)

    Roth, J.

    1981-01-01

    The g-g nucleus 74 Kr was studied by means of the reaction 58 Ni ( 19 F, p2n#betta#) 74 Kr. In order to make gamma spectroscopic studies at neutron deficient nuclei like 74 Kr a neutron multiplicity measurement technique was developed. Beside #betta# single spectra, #betta# excitation functions, #betta#-#betta# coincidences, #betta# angular distributions, and lifetime measurements by means of this technique all measurements in coincidence with up to two neutrons were taken up. From these measurement data an extended term scheme with 17 newly found excited states could be extracted. To all levels spins and parities could be assigned. From the four energetically lowest levels of the yrast cascade the mean lifetimes could be determined. A double backbending in the sequence of the yrast cascade was interpreted as crossing of the g 9/2 bands. The irregularities in the lower part of the yrast band correspond to the shape consistence picture. The results were considered in connection with the systematics of the even krypton isotopes and compared with a two-quasiparticle-plas-rotor model calculation. (HSI)

  6. Shock absorber

    International Nuclear Information System (INIS)

    Nemeth, J.D.

    1981-01-01

    A shock absorber for the support of piping and components in a nuclear power plant is described. It combines a high degree of stiffness under sudden shocks, e.g. seismic disturbances, with the ability to allow for thermal expansion without resistance when so required. (JIW)

  7. Fluence-to-absorbed-dose conversion coefficients for neutron beams from 0.001 eV to 100 GeV calculated for a set of pregnant female and fetus models

    International Nuclear Information System (INIS)

    Taranenko, Valery; Xu, X George

    2008-01-01

    Protection of fetuses against external neutron exposure is an important task. This paper reports a set of absorbed dose conversion coefficients for fetal and maternal organs for external neutron beams using the RPI-P pregnant female models and the MCNPX code. The newly developed pregnant female models represent an adult female with a fetus including its brain and skeleton at the end of each trimester. The organ masses were adjusted to match the reference values within 1%. For the 3 mm cubic voxel size, the models consist of 10-15 million voxels for 35 organs. External monoenergetic neutron beams of six standard configurations (AP, PA, LLAT, RLAT, ROT and ISO) and source energies 0.001 eV-100 GeV were considered. The results are compared with previous data that are based on simplified anatomical models. The differences in dose depend on source geometry, energy and gestation periods: from 20% up to 140% for the whole fetus, and up to 100% for the fetal brain. Anatomical differences are primarily responsible for the discrepancies in the organ doses. For the first time, the dependence of mother organ doses upon anatomical changes during pregnancy was studied. A maximum of 220% increase in dose was observed for the placenta in the nine months model compared to three months, whereas dose to the pancreas, small and large intestines decreases by 60% for the AP source for the same models. Tabulated dose conversion coefficients for the fetus and 27 maternal organs are provided

  8. Measurement of vortex motion in a type-II superconductor: A novel use of the neutron spin-echo technique

    Science.gov (United States)

    Forgan; Kealey; Johnson; Pautrat; Simon; Lee; Aegerter; Cubitt; Farago; Schleger

    2000-10-16

    We have used the neutron spin-echo technique to measure the small energy change of neutrons which are diffracted by a moving vortex lattice in a low-pinning Nb-Ta superconducting sample. A transport current was passed in the mixed state to cause flux line movement. In the case of uniform motion, the flux velocity v(L) was given as expected by the values of electric and magnetic fields, via E = -v(L)wedgeB. We show that with a nonuniformly moving vortex lattice, one can measure the dispersion of the velocities, opening up new possibilities for investigating moving vortex lines.

  9. Experimental techniques and theoretical models for the study of integral 14 MeV neutron cross sections

    International Nuclear Information System (INIS)

    Csikai, J.

    1981-01-01

    Owing to technical reasons, most of the data for fast neutron-induced reactions were measured at 14 MeV and the free parameters in nuclear reaction models have been determined at this energy. The discrepancies between experiment and theory are often due to the unmeasured or unreliable experimental data; therefore, it is important to survey the present techniques used for the measurement of total, elastic, nonelastic and partial nonelastic [(n,xn); (n,x charged); (n,f); (n,γ)] cross sections for 14 MeV neutrons. Systematics in the data as well as theoretical and semi-empirical models are also outlined. (author)

  10. Extended automated separation techniques in destructive neutron activation analysis; application to various biological materials, including human tissues and blood

    International Nuclear Information System (INIS)

    Tjioe, P.S.; Goeij, J.J.M. de; Houtman, J.P.W.

    1976-09-01

    Neutron activation analysis may be performed as a multi-element and low-level technique for many important trace elements in biological materials, provided that post-irradiation chemical separations are applied. This paper describes a chemical separation consisting of automated procedures for destruction, distillation, and anion-chromatography. The system developed enables the determination of 14 trace elements in biological materials, viz. antimony, arsenic, bromine, cadmium, chromium, cobalt, copper, gold, iron, mercury, molybdenum, nickel, selenium, and zinc. The aspects of sample preparation, neutron irradiation, gamma-spectrum evaluation, and blank-value contribution are also discussed

  11. Segmental dynamics in polymer melts by relaxation techniques and quasielastic neutron scattering

    Science.gov (United States)

    Colmenero, J.

    1993-01-01

    The dynamics of the segmental α-relaxation in three different polymeric systems, poly(vinyl methy ether) (PVME), poly(vinyl chloride) (PVC) and poly(bisphenol A, 2-hydroxypropylether) (PH) has been studied by means of relaxation techniques and quasielastic neutron scattering (backscattering spectrometers IN10 and IN13 at the ILL-Grenoble). By using these techniques we have covered a wide timescale ranging from mesoscopic to macroscopic times (10-10-101s). For analyzing the experimental data we have developed a phenomenological procedure in the frequency domain based on the Havriliak-Negami relaxation function which in fact implies a Kohlrausch-Williams-Watts relaxation function in the time domain. The results obtained indicate that the dynamics of the α-relaxation in a wide timescale shows a clear non-Debye behaviour. The shape of the relaxation function is found to be similar for the different techniques used and independent of temperature and momentum transfer (Q). Moreover the characteristic relaxation times deduced from the fitting of the experimental data can also be described using only one Vogel-Fulcher functional form. Besides we found that the Q-dependence of the relaxation times obtained by QENS is given by a power law, τ(Q) propto Q-n (n > 2) n being dependent on the system, and that the Q-behaviour and the non-Debye behaviour are directly correlated. We discuss this correlation taking into account several data of the dynamics of the α-relaxation previously reported in the literature. We also outline a possible scenario for explaining this empirical correlation.

  12. Neutron activation analysis technique and X-ray fluorescence in bovine liver sample

    International Nuclear Information System (INIS)

    Maihara, V.A.; Favaro, D.I.T.; Vasconcellos, M.B.A.; Sato, I.M.; Salvador, V.L.

    2002-01-01

    Many analytical techniques have been used in food and diet analysis in order to determine a great number of nutritional elements, ranging from percentage to ng g -1 , with high sensitivity and accuracy. Instrumental Neutron activation Analysis (INAA) has been employed to certificate many trace elements in biological reference materials. More recently, the X-Ray Fluorescence (FRX-WD) has been also used to determine some essential elements in food samples. The INAA has been applied in nutrition studies in our laboratory at IPEN since the 80 s. For the development of analytical methodologies the use of the reference materials with the same characteristics of the sample analyzed is essential. Several Brazilian laboratories do not have conditions to use these materials due their high cost.In this paper preliminary results of commercial bovine liver sample analyses obtained by INAA and WD-XRF methods are presented. This sample was prepared to be a Brazilian candidate of reference material for a group of laboratories participating in a research project sponsored by FAPESP. The concentrations of some elements like Cl, K, Na, P, S and trace elements Br, Ca, Co, Cu, Fe, Mg, Mn, Mo, Rb, Se and Zn were determined by INAA and WD-XFR. For validation methodology of both techniques, NIST SRM 1577b Bovine Liver reference material was analyzed and the detection limits were calculated. The concentrations of elements determined by both analytical techniques were compared by using the Student's t-test and for Cl, Cu, Fe, K, Mg, Na, Rn and Zn the results do show no statistical difference for 95% significance level. (author)

  13. An application of reactor noise techniques to neutron transport problems in a random medium

    International Nuclear Information System (INIS)

    Sahni, D.C.

    1989-01-01

    Neutron transport problems in a random medium are considered by defining a joint Markov process describing the fluctuations of one neutron population and the random changes in the medium. Backward Chapman-Kolmogorov equations are derived which yield an adjoint transport equation for the average neutron density. It is shown that this average density also satisfied the direct transport equation as given by the phenomenological model. (author)

  14. 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)

  15. Research within the coordinated programme on neutron scattering techniques in applied research

    International Nuclear Information System (INIS)

    Satya Murthy, N.S.

    1982-02-01

    This paper reviews developments of neutron scattering studies at Bhabha Atomic Research Centre (BARC) over the past two decades through utilisation of Apsara and Circus reactor facilities. Salient results in neutron crystallography, magnetic diffraction and inelastic neutron scattering will be presented highlighting progressive involvement in more and more complex studies. The growth of non-neutronic activities as a natural outcome of overall necessity and interest of investigators will be indicated. A description of facilities planned at R5 and the nature of studies that are likely to be taken up at R5 will be briefly discussed. (author)

  16. Study of two-zone reactor system using a pulsed neutron technique

    Energy Technology Data Exchange (ETDEWEB)

    Shishin, B P; Platovskikh, Yu A; Didejkin, T S

    1977-05-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t..-->..infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations.

  17. Elemental analysis of water and soil environmental samples in Tabuk area by neutron capture gamma-ray spectroscopy techniques

    International Nuclear Information System (INIS)

    Al-Aseery, Sh.M.; Alamoudi, Z.; Hassan, A.M.

    2006-01-01

    The prompt and delayed gamma-rays due to neutron capture in the nuclei of the constituent elements of three soil samples and one drinking water sample have been measured. The 252 Cf and 226 Ra/Be isotopic neutron sources are used for neutron irradiation. Also, the hyper pure germanium detection system is used. The soil samples were from Astra, Tadco and El-Gammaz farms, while the water sample was taken from Tabuk city. In case of prompt gamma-ray analysis, a total of 16 elements were identified and the concentration percentage values by weight were calculated for: C, Na, Mg, Al, Si, S, Cl,, Ca, Ti, Cr, Mn, Fe, Co, Zn, Sr ad Pb elements. A comparative study between the results obtained in this work and the results obtained by ICP-MS and EDX-Ray techniques for the same samples is given

  18. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  19. Verification of threshold activation detection (TAD) technique in prompt fission neutron detection using scintillators containing 19F

    Science.gov (United States)

    Sibczynski, P.; Kownacki, J.; Moszyński, M.; Iwanowska-Hanke, J.; Syntfeld-Każuch, A.; Gójska, A.; Gierlik, M.; Kaźmierczak, Ł.; Jakubowska, E.; Kędzierski, G.; Kujawiński, Ł.; Wojnarowicz, J.; Carrel, F.; Ledieu, M.; Lainé, F.

    2015-09-01

    In the present study ⌀ 5''× 3'' and ⌀ 2''× 2'' EJ-313 liquid fluorocarbon as well as ⌀ 2'' × 3'' BaF2 scintillators were exposed to neutrons from a 252Cf neutron source and a Sodern Genie 16GT deuterium-tritium (D+T) neutron generator. The scintillators responses to β- particles with maximum endpoint energy of 10.4 MeV from the n+19F reactions were studied. Response of a ⌀ 5'' × 3'' BC-408 plastic scintillator was also studied as a reference. The β- particles are the products of interaction of fast neutrons with 19F which is a component of the EJ-313 and BaF2 scintillators. The method of fast neutron detection via fluorine activation is already known as Threshold Activation Detection (TAD) and was proposed for photofission prompt neutron detection from fissionable and Special Nuclear Materials (SNM) in the field of Homeland Security and Border Monitoring. Measurements of the number of counts between 6.0 and 10.5 MeV with a 252Cf source showed that the relative neutron detection efficiency ratio, defined as epsilonBaF2 / epsilonEJ-313-5'', is 32.0% ± 2.3% and 44.6% ± 3.4% for front-on and side-on orientation of the BaF2, respectively. Moreover, the ⌀ 5'' EJ-313 and side-on oriented BaF2 were also exposed to neutrons from the D+T neutron generator, and the relative efficiency epsilonBaF2 / epsilonEJ-313-5'' was estimated to be 39.3%. Measurements of prompt photofission neutrons with the BaF2 detector by means of data acquisition after irradiation (out-of-beam) of nuclear material and between the beam pulses (beam-off) techniques were also conducted on the 9 MeV LINAC of the SAPHIR facility.

  20. 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)

  1. Calculation of absorbed fractions to human skeletal tissues due to alpha particles using the Monte Carlo and 3-d chord-based transport techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, J.G. [Institute of Radiation Protection and Dosimetry, Av. Salvador Allende s/n, Recreio, Rio de Janeiro, CEP 22780-160 (Brazil); Watchman, C.J. [Department of Radiation Oncology, University of Arizona, Tucson, AZ, 85721 (United States); Bolch, W.E. [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL, 32611 (United States); Department of Biomedical Engineering, University of Florida, Gainesville, FL 32611 (United States)

    2007-07-01

    Absorbed fraction (AF) calculations to the human skeletal tissues due to alpha particles are of interest to the internal dosimetry of occupationally exposed workers and members of the public. The transport of alpha particles through the skeletal tissue is complicated by the detailed and complex microscopic histology of the skeleton. In this study, both Monte Carlo and chord-based techniques were applied to the transport of alpha particles through 3-D micro-CT images of the skeletal microstructure of trabecular spongiosa. The Monte Carlo program used was 'Visual Monte Carlo-VMC'. VMC simulates the emission of the alpha particles and their subsequent energy deposition track. The second method applied to alpha transport is the chord-based technique, which randomly generates chord lengths across bone trabeculae and the marrow cavities via alternate and uniform sampling of their cumulative density functions. This paper compares the AF of energy to two radiosensitive skeletal tissues, active marrow and shallow active marrow, obtained with these two techniques. (authors)

  2. Experimental approaches for distribution and behavior of water in PEMFC under flow direction and differential pressure using neutron imaging technique

    International Nuclear Information System (INIS)

    Kim, TaeJoo; Kim, JongRok; Sim, CheulMuu; Lee, SeungWook; Kaviany, Massound; Son, SangYoung; Kim, MooHwan

    2009-01-01

    In this investigation, we prepared a 3-parallel serpentine single PEMFC which has an active area of 25 cm 2 and a flow channel cross section of 1x1 mm. Distribution and transport of water in an operating PEMFC were observed by varying the flow directions (co-current and counter-current) in each channel and the differential pressures (100, 200, 300 kPa) applied between the anode and cathode channels. This investigation was performed at the neutron imaging facility at the NIST of which the collimation ratio and neutron fluence rate are 600, 7.2x10 6 n/s/cm 2 , respectively. Neutron image was continuously recorded by an amorphous silicon flat panel detector every 1 s during the operation of the fuel cell. It has been observed that the differential pressure affects the total amount of water produced while the flow direction affects the spatial distribution of water when the neutron images were analyzed for several different operating conditions. More specifically, the amount of water production in the fuel cell increased as the partial pressure increases at a given current density and the water production was more uniform for the counter current than the co-current case. It is shown that the neutron imaging technique is a powerful tool to visualize the PEMFC. The information on the water distribution and behavior at an operating PEMFC helps improve the efficiency of PEMFC.

  3. Pulsed and monoenergetic beams for neutron cross-section measurements using activation and scattering techniques at Triangle Universities Nuclear Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Hutcheson, A. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States)]. E-mail: hutch@tunl.duke.edu; Angell, C.T. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Becker, J.A. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Boswell, M. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Crowell, A.S. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Dashdorj, D. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Fallin, B. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Fotiades, N. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Howell, C.R.; Karwowski, H.J.; Kelley, J.H.; Kiser, M. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Macri, R.A. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Nelson, R.O. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Pedroni, R.S. [NC A and T State University, 1601 East Market Street, Greensboro, NC 27411 (United States); Tonchev, A.P.; Tornow, W. [Triangle Universities Nuclear Laboratory, P.O. Box 90308, Durham, NC 27708 (United States); Vieira, D.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Weisel, G.J. [Penn State Altoona, 3000 Ivyside Park, Altoona, PA 16601 (United States); Wilhelmy, J.B. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2007-08-15

    In support of the Stewardship Science Academic Alliances initiative, an experimental program has been developed at Triangle Universities Nuclear Laboratory (TUNL) to measure (n,xn) cross-sections with both in-beam and activation techniques with the goal of improving the partial cross-section database for the NNSA Stockpile Stewardship Program. First experimental efforts include excitation function measurements on {sup 235,238}U and {sup 241}Am using pulsed and monoenergetic neutron beams with E {sub n} = 5-15 MeV. Neutron-induced partial cross-sections were measured by detecting prompt {gamma} rays from the residual nuclei using various combinations of clover and planar HPGe detectors in the TUNL shielded neutron source area. Complimentary activation measurements using DC neutron beams have also been performed in open geometry in our second target area. The neutron-induced activities were measured in the TUNL low-background counting area. In this presentation, we include detailed information about the irradiation procedures and facilities and preliminary data on first measurements using this capability.

  4. Study of the Condensed Matter Dynamics by the Deep Inelastic Neutron Technique

    International Nuclear Information System (INIS)

    Blostein, Juan Jeronimo

    2004-01-01

    In this work we describe the motivations that originated the eVS technique, and we explain why this technique is particularly useful to study the dynamics of nuclei in different systems.We expose the fundamentals of the convolution formalism usually employed for the data treatment of this technique, which is based on the assumption that the energy of the detected neutrons is fixed by the resonant filter. In this thesis it is presented for the first time the exact formalism that describes the experimentally obtained spectra for thin samples, which considers all pairs of initial and final energies compatible with the flight paths, in clear contrast with the mentioned hypothesis of fixed final energy .In the light of the exact formalism, the analysis presented in the this thesis reveals flaws in the convolution formalism employed for the experimental data treatment.This analysis shows that the deficient descriptions of the different peaks observed in the time of flight scale produce errors in the magnitudes usually informed as results of an eVS experiment, i.e.: intensity of isolated peaks, intensities of peaks not totally resolved, effective temperatures and momentum distributions.To expose the basic reason of such deficiencies, an the exact formalism is employed to analyze the final energy distributions responsible of the observed intensities in the different time channels.We conclude that for each time channel, it is not a well-defined energy, but a whole energy distribution the responsible to produce the observed spectrum.We show that such energy distributions are not determined only by the filter's resonance, but depend strongly on the time of flight due to kinematic and dynamic reasons, not taken into account in the usual data treatment.We analyze examples where the use of the convolution approximation leads to erroneous results, and we show them to have a similar behavior than claimed anomalous results published in several papers, which ascribe them to new

  5. Metallic Concentration in Karnaphuly Estuary Sediment Using Neutron Activation Analysis Technique

    International Nuclear Information System (INIS)

    Rahman, A.K.M.R.; Chakroborty, S.R.; Roy, P.K.; Dev, A.K.

    2013-01-01

    Thousands of tonnes of liquid waste, dirt and garbage from different mills and factories from Kaptai to Chittagong are dumped unhindered into the Karnaphuli River. Pollutants enter the river directly from urban sewerage and industrial waste discharges. Ships pollute waterways in many ways. All theses can contribute to the heavy and other metals of the water bodies. To find out concentration of different metallic elements and make a comparison with the world reference values twenty one surface sediment samples were collected from the different locations of Karnaphuly River near Chittagong city drainage outlet. Neutron Activation Analysis (NAA) technique was used for detecting the concentration qualitatively and quantitatively using the 3 MW TRIGA MARK II research reactor of Atomic Energy Research Establishment, Savar, Dhaka. Pearson correlation among the elements obtained using SPSS software. Three transitional metals Fe, Cr and Co, two alkali metals Rb and Na, two lanthanides La and Ce, and two metalloids As and Sb were determined. The concentrations of elements were compared with the world reference values. Some elements are found in elevated level. (author)

  6. Investigation for Water Propagation at PEMFC with Single Channel by Neutron Imaging Technique

    International Nuclear Information System (INIS)

    Kim, Tae Joo; Sim, Cheul Muu; Kim, Jong Rok; Kim, Moo Hwan

    2008-01-01

    Effective water management increases performance and durability of the Polymer Electrolyte Membrane Fuel cell(PEMFC). The membrane in PEMFC must be sufficiently hydrated because its conductivity relies primarily on the humidity state of the membrane. Since water is generated as a by-product when the fuel cell is generating power, this water source can be said to be a 'disturbance' to any water management system, which is trying to maintain proper humidity level without flooding. Since water is generated throughout the active area, the downstream area can be flooded even when the upstream area is under-saturated. This creates a challenging environment for water management, which adversely affects the efficiency and reliability in the operation of the PEMFC. Although there are many researches for the water management, their interests are limited on the performance. However, the fundamental information of water propagation characteristics is needed to make a scheme for water management. In this study, we used specially designed PEMFC with only single channel, and the water propagation was investigated according to the channel location by neutron imaging technique

  7. Characterisation of air particulate matter in Klang Valley by neutron activation analysis technique

    International Nuclear Information System (INIS)

    Mohd Suhaimi Hamzah; Shamsiah Abd Rahman; Mohd Khalid Matori; Abd Khalik Wood

    2000-01-01

    Air particulate matter is known to affect human health, impairs visibility and can cause climate change. Study on air particulate matter in term of particle size and chemical contents is very important to indicate the quality of air in a sampling area. Information on concentration of important constituents in air particles can be used to identify some of emission sources which contribute to the pollution problem. The data collected may also be, used as a basis to design a strategy in order to overcome the air pollution problem in the area. The study involved sampling of air dust at two stations, one in Bangi and the other in Kuala Lumpur using Gent Stack Sampler units. Each sampler capable of collecting air particle sizes smaller than 2.5 micron (PM 2.5) and between 2.5 - O micron on two different filters simultaneously. The filters were measured for their mass, elemental carbon and elemental concentrations using analytical equipment or techniques including reflectometer and Neutron Activation Analysis. The results of analysis on samples collected in 1997-1998 are discussed. (author)

  8. Characterization of Nanocellulose Using Small-Angle Neutron, X-ray, and Dynamic Light Scattering Techniques.

    Science.gov (United States)

    Mao, Yimin; Liu, Kai; Zhan, Chengbo; Geng, Lihong; Chu, Benjamin; Hsiao, Benjamin S

    2017-02-16

    Nanocellulose extracted from wood pulps using TEMPO (2,2,6,6-tetramethylpiperidine-1-oxyl radical)-mediated oxidation and sulfuric acid hydrolysis methods was characterized by small-angle neutron scattering (SANS), small-angle X-ray scattering (SAXS), and dynamic light scattering (DLS) techniques. The dimensions of this nanocellulose (TEMPO-oxidized cellulose nanofiber (TOCN) and sulfuric acid hydrolyzed cellulose nanocrystal (SACN)) revealed by the different scattering methods were compared with those characterized by transmission electron microscopy (TEM). The SANS and SAXS data were analyzed using a parallelepiped-based form factor. The width and thickness of the nanocellulose cross section were ∼8 and ∼2 nm for TOCN and ∼20 and ∼3 nm for SACN, respectively, where the fitting results from SANS and SAXS profiles were consistent with each other. DLS was carried out under both the V V mode with the polarizer and analyzer parallel to each other and the H V mode having them perpendicular to each other. Using rotational and translational diffusion coefficients obtained under the H V mode yielded a nanocellulose length qualitatively consistent with that observed by TEM, whereas the length derived by the translational diffusion coefficient under the V V mode appeared to be overestimated.

  9. Variance reduction techniques for 14 MeV neutron streaming problem in rectangular annular bent duct

    Energy Technology Data Exchange (ETDEWEB)

    Ueki, Kotaro [Ship Research Inst., Mitaka, Tokyo (Japan)

    1998-03-01

    Monte Carlo method is the powerful technique for solving wide range of radiation transport problems. Its features are that it can solve the Boltzmann`s transport equation almost without approximation, and that the complexity of the systems to be treated rarely becomes a problem. However, the Monte Carlo calculation is always accompanied by statistical errors called variance. In shielding calculation, standard deviation or fractional standard deviation (FSD) is used frequently. The expression of the FSD is shown. Radiation shielding problems are roughly divided into transmission through deep layer and streaming problem. In the streaming problem, the large difference in the weight depending on the history of particles makes the FSD of Monte Carlo calculation worse. The streaming experiment in the 14 MeV neutron rectangular annular bent duct, which is the typical streaming bench mark experiment carried out of the OKTAVIAN of Osaka University, was analyzed by MCNP 4B, and the reduction of variance or FSD was attempted. The experimental system is shown. The analysis model by MCNP 4B, the input data and the results of analysis are reported, and the comparison with the experimental results was examined. (K.I.)

  10. GENERALISATION OF RADIATOR DESIGN TECHNIQUES FOR PERSONAL NEUTRON DOSEMETERS BY UNFOLDING METHOD.

    Science.gov (United States)

    Oda, K; Nakayama, T; Umetani, K; Kajihara, M; Yamauchi, T

    2016-09-01

    A novel technique for designing a radiator suitable for personal neutron dosemeter based on plastic track detector was discussed. A multi-layer structure has been proposed in the previous report, where the thicknesses of plural polyethylene (PE) layers and insensitive ones were determined by iterative calculations of double integral. In order to arrange this procedure and make it more systematic, unfolding calculation has been employed to estimate an ideal radiator containing an arbitrary hydrogen concentration. In the second step, realistic materials replaced it with consideration of minimisation of the layer number and commercial availability. A radiator consisting of three layers of PE, Upilex and Kapton sheets was finally designed, for which a deviation in the energy dependence between 0.1 and 20 MeV could be controlled within 18 %. An applicability of fluorescent nuclear track detector element has also been discussed. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  11. Study of two-zone reactor system using a pulsed neutron technique

    International Nuclear Information System (INIS)

    Shishin, B.P.; Platovskikh, Yu.A.; Didejkin, T.S.

    1977-01-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t→infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations. The experiments show that the observed neutron flux decrease in the core is governed by an exponent exp(-Λ 1 t), whereas in the reflector by a sum of two exponents exp(-Λ 1 t) and exp(-Λ 2 t). The eigenvalue Λ 1 reflects multiplying properties of the reactor, and Λ 2 is determined by the reflector absorption cross section

  12. Nondestructive determination of plutonium mass in spent fuel: preliminary modeling results using the passive neutron Albedo reactivity technique

    International Nuclear Information System (INIS)

    Evans, Louise G.; Tobin, Stephen J.; Schear, Melissa A.; Menlove, Howard O.; Lee, Sang Y.; Swinhoe, Martyn T.

    2009-01-01

    There are a variety of motivations for quantifying plutonium (Pu) in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capability of the International Atomic Energy Agency (LAEA) to safeguard nuclear facilities, quantifying shipper/receiver difference, determining the input accountability value at pyrochemical processing facilities, providing quantitative input to burnup credit and final safeguards measurements at a long-term repository. In order to determine Pu mass in spent fuel assemblies, thirteen NDA techniques were identified that provide information about the composition of an assembly. A key motivation of the present research is the realization that none of these techniques, in isolation, is capable of both (1) quantifying the Pu mass of an assembly and (2) detecting the diversion of a significant number of rods. It is therefore anticipated that a combination of techniques will be required. A 5 year effort funded by the Next Generation Safeguards Initiative (NGSI) of the U.S. DOE was recently started in pursuit of these goals. The first two years involves researching all thirteen techniques using Monte Carlo modeling while the final three years involves fabricating hardware and measuring spent fuel. Here, we present the work in two main parts: (1) an overview of this NGSI effort describing the motivations and approach being taken; (2) The preliminary results for one of the NDA techniques - Passive Neutron Albedo Reactivity (PNAR). The PNAR technique functions by using the intrinsic neutron emission of the fuel (primarily from the spontaneous fission of curium) to self-interrogate any fissile material present. Two separate measurements of the spent fuel are made, both with and without cadmium (Cd) present. The ratios of the Singles, Doubles and Triples count rates obtained in each case are analyzed; known as the Cd ratio. The primary differences between the two measurements are the neutron energy spectrum

  13. Basic of Neutron NDA

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-15

    The objectives of this presentation are to introduce the basic physics of neutron production, interactions and detection; identify the processes that generate neutrons; explain the most common neutron mechanism, spontaneous and induced fission and (a,n) reactions; describe the properties of neutron from different sources; recognize advantages of neutron measurements techniques; recognize common neutrons interactions; explain neutron cross section measurements; describe the fundamental of 3He detector function and designs; and differentiate between passive and active assay techniques.

  14. Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

    International Nuclear Information System (INIS)

    Jesus Barbosa, S. de.

    1987-01-01

    The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.) [pt

  15. Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on 235U and 239Pu using the double time-of-flight technique

    International Nuclear Information System (INIS)

    Noda, S.; Haight, R. C.; Nelson, R. O.; Devlin, M.; O'Donnell, J. M.; Chatillon, A.; Granier, T.; Belier, G.; Taieb, J.; Kawano, T.; Talou, P.

    2011-01-01

    Prompt fission neutron spectra from 235 U and 239 Pu were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg 235 U and 90 mg 239 Pu detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced γ-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

  16. 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.)

  17. Amino acids analysis using grouping and parceling of neutrons cross sections techniques

    International Nuclear Information System (INIS)

    Voi, Dante Luiz Voi; Rocha, Helio Fenandes da

    2002-01-01

    Amino acids used in parenteral administration in hospital patients with special importance in nutritional applications were analyzed to compare with the manufactory data. Individual amino acid samples of phenylalanine, cysteine, methionine, tyrosine and threonine were measured with the neutron crystal spectrometer installed at the J-9 irradiation channel of the 1 kW Argonaut Reactor of the Instituto de Engenharia Nuclear (IEN). Gold and D 2 O high purity samples were used for the experimental system calibration. Neutron cross section values were calculated from chemical composition, conformation and molecular structure analysis of the materials. Literature data were manipulated by parceling and grouping neutron cross sections. (author)

  18. A comparative examination of several techniques for the routine determination of mercury in biological samples by neutron activation analysis

    International Nuclear Information System (INIS)

    Faanhof, A.; Das, H.A.

    1978-01-01

    A comparative examination of the most important techniques for the separation of mercury from irradiated biological material was made. Procedures for routine analysis and results for standard materials are given. Activation was performed at a thermal neutron flux of approximately 5x10 12 nxcm -2 xs -1 during ( 3 ) 2 offers a convenient solution to this problem. The variation of the neutron flux with the irradiation position can be measured by the application of thin iron rings as flux monitors. Losses of mercury due to uptake in the wall of the irradiation containers are negligible. The most powerful destruction technique for large samples is that based on a stainless-steel bomb. (T. I.)

  19. Absorbed dose by a CMOS in radiotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Valero L, C. Y.; Guzman G, K. A.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L. C., E-mail: candy_borja@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-10-15

    Absorbed dose by a complementary metal oxide semiconductor (CMOS) circuit as part of a pacemaker, has been estimated using Monte Carlo calculations. For a cancer patient who is a pacemaker carrier, scattered radiation could damage pacemaker CMOS circuits affecting patient's health. Absorbed dose in CMOS circuit due to scattered photons is too small and therefore is not the cause of failures in pacemakers, but neutron calculations shown an absorbed dose that could cause damage in CMOS due to neutron-hydrogen interactions. (Author)

  20. Study on measurement technique contrast of 14 MeV neutron fluence

    International Nuclear Information System (INIS)

    Jiang Li; Hu Jun; Wen Dezhi

    2005-10-01

    The stability and repetition of the associated-particle method to measure DT neutron fluence was tested. The neutron activation iron method was contrasted with the associated-particle method, the preparatory experiment was done. The neutron fluence measured with associated-particle method was contrasted with neutron activation Al method, the Al activated foil was measured with 4πβ (PC)-γ coincidence standard device. The contrast result's standard deviation of the two method was less than the expand uncertainty of the associated-particle method. Therein, the uncertainty of the associated-particle method is 1.6%, the uncertainty of the activation Al method is 1.8%. (authors)

  1. Application of neutron activation analysis technique in elemental determination of lichen samples

    International Nuclear Information System (INIS)

    Djoko Prakoso Dwi Atmodjo; Syukria Kurniawati; Woro Yatu Niken Syahfitri; Nana Suherman; Dadang Supriatna

    2010-01-01

    Lichen is one of the biological materials as pollution monitor that can give information about level, direction, and history of various pollutants in environment. Small sample weights and elemental content of lichens is on the order of ppm, so that its characterization requires advanced analytical techniques that has high sensitivity and capable of analyzing samples with weight of - 25 mg, such as neutron activation analysis. In this research, determination of elements was done in lichen samples obtained from Kiaracondong and Holis areas in Bandung city, to understanding the difference of industrial exposure level on surrounding environment. Samples were irradiated in RSG GA Siwabessy, Serpong, at 15 MW for 1-2 and 60 minutes for short irradiation and long irradiation, respectively. The samples were then counted using HPGe detector with GENIE 2000 software. The level of element in lichen for Kiaracondong area were Co, Cr, Cs, Fe, Mg, Mn, Sb, Sc, and V in the range of 0.55-0.86, 1.47-2.57, 0.87-1.19, 540-1005, 949-1674, 34.91-45.94, 0.08-0.14, 0.16-0.31, and ≤ 2.33 mg/kg, respectively, while for Holis area were 1.04-2.37, 4.41-10.36, 0.41-0.89, 3166-709, 1131-1422, 40.97-72.51, 0.33-0.50, 0.98-2.18, and 5.30-13.05 mg/kg respectively. From these results, it is known that pollution exposure from the semi industrial area Holis provide greater influence than in the semi industrial area Kiaracondong. (author)

  2. Three dimensional measurements of absorbed dose in BNCT by Fricke-gel imaging

    International Nuclear Information System (INIS)

    Gambarini, G.; Agosteo, S.; Marchesi, P.; Nava, E.; Palazzi, P.; Pecci, A.; Rosa, R.; Rosi, G.; Tinti, R.

    2001-01-01

    A method has been studied for absorbed dose imaging and profiling in a phantom exposed to thermal or epithermal neutron fields, also discriminating between various contributions to the absorbed dose. The proposed technique is based on optical imaging of FriXy-gel phantoms, which are proper tissue-equivalent phantoms acting as continuous dosimeters. Convenient modifications in phantom composition allow, from differential measurements, the discrimination of various contributions to the absorbed dose. The dosimetry technique is based on a chemical dosimeter incorporated in a tissue-equivalent gel (Agarose). The chemical dosimeter is a ferrous sulphate solution (which is the main component of the standard Fricke dosimeter) added with a metal ion indicator (Xylenol Orange). The absorbed dose is measured by analysing the variation of gel optical absorption in the visible spectrum, imaged by means of a CCD camera provided with a suitable filter. The technique validity has been tested by irradiating and analysing phantoms in the thermal facility of the fast research reactor TAPIRO (ENEA, Casaccia, Italy). In a cylindrical phantom simulating a head, we have imaged the therapy dose from thermal neutron reactions with 10 B and the dose in healthy tissue not containing boron. In tissue without boron, we have discriminated between the two main contributions to the absorbed dose, which comes from the 1 H(n,γ) 2 H and 14 N(n,p) 14 C reactions. The comparison with the results of other experimental techniques and of simulations reveals that the technique is very promising. A method for the discrimination of fast neutron contribution to the absorbed dose, still in an experimental stage, is proposed too. (author)

  3. Utilization of RP-10 reactor for neutron therapy

    International Nuclear Information System (INIS)

    Paucar, R.; Nieto, M.; Parreno, F.; Vela, M.; Pozo, Z.

    1997-01-01

    In the Nuclear Energy Peruvian Institute, IPEN, a research area has established of Neutron Radiotherapy, know as NCT. This research joins the physics of particles (Neutrons and photons) and Medical Physics, and this one is an applied investigation where in considering the construction of a treatment hall in Huarangal (Peru) Reactor's irradiation facility, it can treat patients with brain tumors. In Neutron Therapy (NCT), it tries to use neutrons to destroy tumor cells where other therapeutic techniques are not effective. This process consist on to incise a neutrons beam of adequate characteristics over the tumor area of the patient. The neutrons used are of thermal energy and therefore irradiations are developed in experimental reactors. For this one, it is used horizontal channels prepared suitably. Before the irradiation, it is injected to the patient a substance which is absorbed by tumoral tissue. The substance components will be B-10, nuclide with an absorption cross section high to thermal neutrons (3837 b). The B-10 irradiate with thermal neutrons produce alpha particles of short reach (10 μm. on soft tissue) and with LET values (lineal energy transference) very high. The result is a cell preferential destruction which have absorbed the substance and it's next neighbors, like the cell size is 10 μm. This process as know as Boron Neutron Capture Therapy (BNCT). This work describes Peruvian RP-10 reactor and recently efforts to assess the design and feasibility of the medical neutron irradiation facility for NCT. (author). 22 refs., 6 tabs

  4. Development of accurate techniques for controlling polarization of a long wavelength neutron beam in very low magnetic fields. I

    International Nuclear Information System (INIS)

    Kawai, Takeshi; Ebisawa, Toru; Tasaki, Seiji; Akiyoshi, Tsunekazu; Eguchi, Yoshiaki; Hino, Masahiro; Achiwa, Norio.

    1995-01-01

    The purpose of our study is to develop accurate techniques for controlling polarization of a long wavelength neutron beam and to make a thin-film dynamical spin-flip device operated in magnetizing fields less than 100 gauss and in a shorter switching time up to 20 kHz. The device would work as a chopper for a polarized neutron beam and as a magnetic switching device for a multilayer neutron interferometer. We have started to develop multilayer polarizing mirrors functioning under magnetizing fields less than 100 gauss. The multilayers of Permalloy-Ge and Fe-Ge have been produced using the evaporation method under magnetizing fields of about 100 gauss parallel to the Si-wafer substrate surface. The hysteresis loop for in-plane magnetization of the multilayers were measured to discuss their feasibilities for the polarizing device functioning under very low magnetizing fields. The polarizing efficiencies of Fe-Ge and Permalloy-Ge multilayers were 95 % and 91 % with reflectivities of 50 % and 66 % respectively under magnetizing fields of 80 gauss. The report also discusses problems in applying these multilayer polarizing mirrors to ultracold neutrons. (author)

  5. Applications of Monte Carlo technique in the detection of explosives, narcotics and fissile material using neutron sources

    International Nuclear Information System (INIS)

    Sinha, Amar; Kashyap, Yogesh; Roy, Tushar; Agrawal, Ashish; Sarkar, P.S.; Shukla, Mayank

    2009-01-01

    The problem of illicit trafficking of explosives, narcotics or fissile materials represents a real challenge to civil security. Neutron based detection systems are being actively explored worldwide as a confirmatory tool for applications in the detection of explosives either hidden inside a vehicle or a cargo container or buried inside soil. The development of a system and its experimental testing is a tedious process and to develop such a system each experimental condition needs to be theoretically simulated. Monte Carlo based methods are used to find an optimized design for such detection system. In order to design such systems, it is necessary to optimize source and detector system for each specific application. The present paper deals with such optimization studies using Monte Carlo technique for tagged neutron based system for explosives and narcotics detection hidden in a cargo and landmine detection using backscatter neutrons. We will also discuss some simulation studies on detection of fissile material and photo-neutron source design for applications on cargo scanning. (author)

  6. Standard Test Method for Oxygen Content Using a 14-MeV Neutron Activation and Direct-Counting Technique

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method covers the measurement of oxygen concentration in almost any matrix by using a 14-MeV neutron activation and direct-counting technique. Essentially, the same system may be used to determine oxygen concentrations ranging from over 50 % to about 10 g/g, or less, depending on the sample size and available 14-MeV neutron fluence rates. Note 1 - The range of analysis may be extended by using higher neutron fluence rates, larger samples, and higher counting efficiency detectors. 1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size, shape, and macroscopic density during irradiation and counting to a standard sample of known oxygen content. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of activation analysis using a neutron generator (1). 1.3 The values stated in either SI or inch-pound units are to be regarded...

  7. Primary Displacement Damage Calculation Induced by Neutron and Ion Using Binary Collision Approximation Techniques (Marlowe Code)

    International Nuclear Information System (INIS)

    Mota, F.; Ortiz, C. J.; Vila, R.

    2012-01-01

    The level of damage expected in future fusion reactors conditions is such that the performance of materials and components under these extreme irradiation conditions is still unknown. Considering this scenario, the study of the effects of energetic neutrons generated in fusion reactors on materials is one of the most important research topics to be carried out during next years. The effects of neutron irradiation on materials involve, from a fundamental point of view, two physical phenomena: i) the displacement of atoms from their equilibrium positions in the lattice, which creates point defects, and ii) the generation of nuclear transmutation reactions that contribute to the formation of impurities inside the material, with He and H as the most important ones. The ratio between the levels of He and H, and the amount of point defects is one of the main parameters to understand the effect of the radiation on materials. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on one hand different kinds of current neutron sources to emulate the fusion irradiation environment are available, as for example - Fission power reactor - Spallation sources - Striping Sources: The objective of the International Fusion Materials Irradiation Facility (IFMIF) will be to provide an intense neutron source with adequate energy spectrum to test the suitability of candidate materials for future nuclear fusion power reactor (DEMO). IFMIF will constitute an essential tool in the international strategy towards the achievement of future fusion reactors. b) on the other hand, as these neutron sources have a number of problems and very strict operating conditions, (e.g. the radiological risks), to emulate the effects of fusion neutron on materials, some other facilities can be used. One example is the Spanish initiative TechnoFusion facility which purpose is to serve as technological support for IFMIF and DEMO. The Material

  8. A technique for determining fast and thermal neutron flux densities in intense high-energy (8-30 MeV) photon fields

    International Nuclear Information System (INIS)

    Price, K.W.; Holeman, G.R.; Nath, R.

    1978-01-01

    A technique for measuring fast and thermal neutron fluxes in intense high-energy photon fields has been developed. Samples of phorphorous pentoxide are exposed to a mixed photon-neutron field. The irradiated samples are then dissolved in distilled water and their activation products are counted in a liquid scintillation spectrometer at 95-97% efficiency. The radioactive decay characteristics of the samples are then analyzed to determine fast and thermal neutron fluxes. Sensitivity of this neutron detector to high energy photons has been measured and found to be small. (author)

  9. Reflection measurements of microwave absorbers

    Science.gov (United States)

    Baker, Dirk E.; van der Neut, Cornelis A.

    1988-12-01

    A swept-frequency interferometer is described for making rapid, real-time assessments of localized inhomogeneities in planar microwave absorber panels. An aperture-matched exponential horn is used to reduce residual reflections in the system to about -37 dB. This residual reflection is adequate for making comparative measurements on planar absorber panels whose reflectivities usually fall in the -15 to -25 dB range. Reflectivity measurements on a variety of planar absorber panels show that multilayer Jaumann absorbers have the greatest inhomogeneity, while honeycomb absorbers generally have excellent homogeneity within a sheet and from sheet to sheet. The test setup is also used to measure the center frequencies of resonant absorbers. With directional couplers and aperture-matched exponential horns, the technique can be easily applied in the standard 2 to 40 GHz waveguide bands.

  10. Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

    International Nuclear Information System (INIS)

    Zhou, T.; Hawari, A. I.; Wehring, B. W.

    2007-01-01

    Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

  11. Determination of boron in aqueous solutions by solid state nuclear track detectors technique, using a filtered neutron beam

    International Nuclear Information System (INIS)

    Moraes, M.A.P.V. de; Pugliesi, R.; Khouri, M.T.F.C.

    1985-11-01

    The solid state nuclear track detectors technique has been used for determination of boron in aqueous solutions, using a filtered neutron beam. The particles tracks from the 10 B(n,α)Li 7 reaction were registered in the CR-39 film, chemically etched in a (30%) KOH solution 70 0 C during 90 minutes. The obtained results showed the usefulness of this technique for boron determination in the ppm range. The inferior detectable limit was 9 ppm. The combined track registration efficiency factor K has been evaluated in the solutions, for the CR-39 detector and its values is K= (4,60 - + 0,06). 10 -4 cm. (Author) [pt

  12. PREFACE: Exploring surfaces and buried interfaces of functional materials by advanced x-ray and neutron techniques Exploring surfaces and buried interfaces of functional materials by advanced x-ray and neutron techniques

    Science.gov (United States)

    Sakurai, Kenji

    2010-12-01

    This special issue is devoted to describing recent applications of x-ray and neutron scattering techniques to the exploration of surfaces and buried interfaces of various functional materials. Unlike many other surface-sensitive methods, these techniques do not require ultra high vacuum, and therefore, a variety of real and complicated surfaces fall within the scope of analysis. It must be particularly emphasized that the techniques are capable of seeing even buried function interfaces as well as the surface. Furthermore, the information, which ranges from the atomic to mesoscopic scale, is highly quantitative and reproducible. The non-destructive nature of the techniques is another important advantage of using x-rays and neutrons, when compared with other atomic-scale analyses. This ensures that the same specimen can be measured by other techniques. Such features are fairly attractive when exploring multilayered materials with nanostructures (dots, tubes, wires, etc), which are finding applications in electronic, magnetic, optical and other devices. The Japan Applied Physics Society has established a group to develop the research field of studying buried function interfaces with x-rays and neutrons. As the methods can be applied to almost all types of materials, from semiconductor and electronic devices to soft materials, participants have fairly different backgrounds but share a common interest in state-of-the-art x-ray and neutron techniques and sophisticated applications. A series of workshops has been organized almost every year since 2001. Some international interactions have been continued intensively, although the community is part of a Japanese society. This special issue does not report the proceedings of the recent workshop, although all the authors are in some way involved in the activities of the above society. Initially, we intended to collect quite long overview papers, including the authors' latest and most important original results, as well as

  13. Hidden explosives detector employing pulsed neutron and x-ray interrogation

    International Nuclear Information System (INIS)

    Schultz, F.J.; Caldwell, J.T.

    1993-01-01

    Methods and systems for the detection of small amounts of modern, highly-explosive nitrogen-based explosives, such as plastic explosives, hidden in airline baggage. Several techniques are employed either individually or combined in a hybrid system. One technique employed in combination is X-ray imaging. Another technique is interrogation with a pulsed neutron source in a two-phase mode of operation to image both nitrogen and oxygen densities. Another technique employed in combination is neutron interrogation to form a hydrogen density image or three-dimensional map. In addition, deliberately-placed neutron-absorbing materials can be detected

  14. Hidden explosives detector employing pulsed neutron and x-ray interrogation

    Science.gov (United States)

    Schultz, Frederick J.; Caldwell, John T.

    1993-01-01

    Methods and systems for the detection of small amounts of modern, highly-explosive nitrogen-based explosives, such as plastic explosives, hidden in airline baggage. Several techniques are employed either individually or combined in a hybrid system. One technique employed in combination is X-ray imaging. Another technique is interrogation with a pulsed neutron source in a two-phase mode of operation to image both nitrogen and oxygen densities. Another technique employed in combination is neutron interrogation to form a hydrogen density image or three-dimensional map. In addition, deliberately-placed neutron-absorbing materials can be detected.

  15. Development and application of some fast neutron dosimetry techniques utilizing plastic track detectors for therapeutic and health physics related applications. Progress report

    International Nuclear Information System (INIS)

    Morgan, K.Z.; Sohrabi, M.

    1975-01-01

    The electrochemical-etch-foil technique continues to look extremely promising for neutron monitoring of personnel and applications in medicine and research. Some of the most important parameters that must be controlled are presented and discussed. (U.S.)

  16. A physical and engineering study on the irradiation techniques in neutron capture therapy aiming for wider application

    International Nuclear Information System (INIS)

    Sakurai, Y.; Ono, K.; Suzuki, M.; Katoh, I.; Miyatake, S.-I.; Yanagie, H.

    2003-01-01

    The solo-irradiation of thermal neutrons has been applied for brain cancer and malignant melanoma in the boron neutron capture therapy (BNCT) at the medical irradiation facility of Kyoto University Reactor (KUR), from the first clinical trial in 1974. In 1997, after the facility remodeling, the application of the mix-irradiation of thermal and epi-thermal neutrons was started, and the depth dose distribution for brain cancer has been improved in some degree. In 2001, the solo-irradiation of epi-thermal neutrons also started. It is specially mentioned that the application to oral cancers started at the same time. The BNCT clinical trial using epi-thermal neutron irradiation at KUR, amounts to twelve as of March 2003. The seven trials; more than a half of the total trials, are for oral cancers. From this fact, we think that the wider application to the other cancers is required for the future prosperity of BNCT. The cancers applied for BNCT in KUR at the present time, are brain cancer, melanoma and oral cancers, as mentioned above. The cancers, expected to be applied in near future, are liver cancer, pancreas cancer, lung cancer, tongue cancer, breast cancer, etc.. Any cancer is almost incurable by the other therapy including the other radiation therapy. In the wider application of BNCT to these cancers, the dose-distribution control suitable to each cancer and/or each part, is important. The introduction of multi-directional and/or multi-divisional irradiation is also needed. Here, a physical and engineering study using two-dimensional transport calculation and three-dimensional Monte-Carlo simulation for the irradiation techniques in BNCT aiming for wider application is reported

  17. Absorber materials in CANDU PHWR's

    International Nuclear Information System (INIS)

    Price, E.G.; Boss, C.R.; Novak, W.Z.; Fong, R.W.L.

    1995-03-01

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in a relatively benign environment of low pressure, low temperature heavy water between neighbouring rows of columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a designed back-fit resolved the problem. (author). 3 refs., 1

  18. The multianalyser system of the three axes neutron spectrometer PUMA: Pilot experiments with the innovative multiplex technique

    Energy Technology Data Exchange (ETDEWEB)

    Sobolev, Oleg; Hoffmann, Ron; Gibhardt, Holger [Institute for Physical Chemistry, Georg-August-University of Göttingen, Tammannstr. 6, D-37077 Göttingen (Germany); Jünke, Norbert [Forschungs-Neutronenquelle Heinz-Maier-Leibnitz, Technical University of Munich, Lichtenbergstr. 1, D-85748 Garching (Germany); Knorr, Andreas; Meyer, Volker [Institute for Physical Chemistry, Georg-August-University of Göttingen, Tammannstr. 6, D-37077 Göttingen (Germany); Eckold, Götz, E-mail: geckold@gwdg.de [Institute for Physical Chemistry, Georg-August-University of Göttingen, Tammannstr. 6, D-37077 Göttingen (Germany)

    2015-02-01

    A new type of multiplex technique for three axes neutron spectrometers has been realized and successfully commissioned at the PUMA spectrometer at FRM II. Consisting of eleven analyser-detector channels which can be configured individually, this technique is especially suitable for kinetic experiments where a single excitation spectrum is recorded as a function of time without the need to move the spectrometer. On a time-scale of seconds an entire spectrum can be recorded thus allowing users to monitor changes during fast kinetic processes in single shot experiments without the need for stroboscopic techniques. Moreover, the multianalyser system provides an efficient and rapid tool for mapping excitations in (Q,ω)-space. The results of pilot experiments demonstrate the performance of this new technique and a user-friendly software is presented which assists users during their experiments.

  19. Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques

    International Nuclear Information System (INIS)

    Ampomah-Amoako, E.

    2008-06-01

    The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)

  20. Determination of boron in water solution by an indirect neutron activation technique from a 241Am/Be source

    International Nuclear Information System (INIS)

    Sales, H.B.

    1981-08-01

    Boron content in water solutions has been analysed by Indirect Activation Technique a twin 241 Am/Be neutron source with a source strength of 9x10 6 n/seg. The boron concentration was inferred from the measurement of the activity induced in a vanadium flux monitor. The vanadium rod was located inside the boron solution in a standart geometrical set up with respect to the neutron source. Boron concentrations in the range of 100 to 1000 ppm were determined with an overall accuracy of about 2% during a total analysis time of about 20 minutes. Eventhough the analysis is not selective for boron yet due the rapid, simple and precise nature, it is proposed for the analysis of boron in the primary coolant circuit of Nuclear Power Plants of PWR type. (Author) [pt

  1. Application of new radiation detection techniques at the Paul Scherrer Institut, especially at the spallation neutron source

    CERN Document Server

    Lehmann, E; Williams, T; Pralong, C

    1999-01-01

    The demands on modern irradiation detection systems are diverse, encompassing spatial resolution, dynamic range, sensitivity and reproducibility. Nevertheless, there are two important new methods which can satisfy most of these demands in several applications: camera based systems and imaging plates. Imaging plates have primarily been used as gamma- and beta-sensitive detectors in biology and medicine, but are now available also as neutron sensitive systems. These methods are ideally suited for applications in neutron radiography because of their high sensitivity, linearity and digital output. Image processing, quantification of the image data and automated pattern recognition can easily be performed using modern software tools. The imaging plate system at PSI is shared between groups in reactor physics, radiation protection, biology, proton therapy and nuclear medicine. The collected experience from these different interests establishes the basis for a most effective application of this technique. The utilis...

  2. Study on bioavailability of zinc for children's diet by using activable isotopic tracer 70Zn and neutron activation analysis techniques

    International Nuclear Information System (INIS)

    Zhang Yangmei; Ni Bangfa; Wang Pingsheng; Tian Weizhi; Cao Lei

    2001-01-01

    Bioavailability of zinc for three groups (low amount of diet zinc, balance amount of diet zinc and high amount of diet zinc) of children's diet is studied by using activable isotopic tracer 70 Zn and neutron activation analysis techniques. The results indicate that the fractional absorption of zinc from balance diet zinc group is the highest, up to 33.9%. A procedure of pre-irradiation concentration zinc for fecal samples using anion exchanger is developed, and the enriched 70 Zn with isotopic abundance of 18.3% is used for tracer. The mass ratios between 70 Zn and 68 Zn or 64 Zn and their contents between natural zinc and enriched zinc are used to calculate the bioavailability of zinc. Instrumental neutron activation analysis of 64 Zn of each original fecal samples and pre-irradiation concentrated zinc samples are used to normalize the chemical yield in order to reduce the uncertainty during the chemical separation procedure

  3. Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

    Energy Technology Data Exchange (ETDEWEB)

    Campo B, X.; Mendez V, R.; Embid S, M. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense 40, 28040 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Sanz G, J., E-mail: xandra.campo@ciemat.es [Universidad Nacional de Educacion a Distancia, Escuela Tecnica Superior de Ingenieros Industriales, C. Juan del Rosal 12, 28040 Madrid (Spain)

    2014-08-15

    Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with {sup 241}Am-Be (185 GBq) and {sup 252}Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the {sup 252}Cf source. This source predominates and the effect of the {sup 241}Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

  4. Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

    International Nuclear Information System (INIS)

    Campo B, X.; Mendez V, R.; Embid S, M.; Vega C, H. R.; Sanz G, J.

    2014-08-01

    Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with 241 Am-Be (185 GBq) and 252 Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the 252 Cf source. This source predominates and the effect of the 241 Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

  5. On the sensitivity evaluation for on-stream elemental analysis using neutron and photoactivation techniques

    International Nuclear Information System (INIS)

    Ivanov, I.N.; Zakharov, E.A.; Kasatkin, V.A.; Kartashev, E.R.; Mashinin, V.A.; Feoktistov, Yu.V.; Tchulkin, V.L.

    1984-01-01

    The paper discusses the data of calculations of neutron and photonactivation effects of moving substances, in particular, bulk materials moving on a conveyor belt and solution streams. The calculated data on minimum detectable concentrations of the various elements are given. The calculation data satisfactorily agree with the experimental data obtained as a result of on-stream neutron activation analysis of solutions containing indium, selenium, aluminium, fluorine and also bulk materials containing aluminium and silicon. The opportunity of applying the bremsstrahlung generated by the electron accelerator for photon activation analysis and sorting gold-bearing ores on the conveyor belt has been analysed

  6. Absorption and activation techniques in measurements of fast-neutron capture cross sections

    International Nuclear Information System (INIS)

    Bergqvist, I.

    1982-01-01

    The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods. The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons causes the main problems

  7. Aborption and activation techniques in measurements of fast-neutron capture cross sections

    International Nuclear Information System (INIS)

    Bergqvist, I.

    1982-01-01

    The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons c causes the main problems. (Author)

  8. Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

    International Nuclear Information System (INIS)

    Kryter, R.C.

    1975-01-01

    Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

  9. Determination of Heavy and Toxic Trace Elements in Sediments of Qarun Lake Using Instrumental Neutron Activation Analysis Technique

    International Nuclear Information System (INIS)

    Hamed, A.; Abd EI-Samad, M.; Soliman, N.F.

    2011-01-01

    An investigation of certain heavy and toxic trace elements in 15 sediment samples from different areas at Qarun Lake was performed by a neutron activation technique using the neutron irradiation facilities of the Second Egyptian Research Reactor (ETRR-2). The samples together with two sets of gold foils (one bare and the other covered with cadmium) and two Standard Reference Materials IAEA (Soil-7) were irradiated at the core of the reactor in two capsules for 8520 seconds, another two capsules each has two sets of gold foils (one bare and the other covered with cadmium) are used to determine the neutron fluxes around the sediment samples. The gamma-rays of nuclides from the irradiated samples were carried out by means of a well calibrated high resolution HPGe detection system. The concentrations of product nuclides containing in the irradiated samples were determined from the peak counting-rates of the prominent gamma-ray lines for the corresponding nuclides using Single External Comparator Method (k 0 -Standardization Method) which called k 0 -NAA technique. The neutron flux ratios (f) in the same radiation sites of the sediment samples bottles were calculated as well as the cadmium ratios (R e d). MS Excel work books were constructed and used in our calculation. The total contents of As , Ba , Ca ,Co, Cr ,Cs, Eu ,Hf ,Hg,Na ,Ru ,Sc ,Se ,Sm ,Sn ,Sr ,Th ,Rb ,Zn ,and Zr in these samples were measured. The objectives .of this study were evaluated to identify the variations of concentration of some elements that have an impact on environmental pollutions. The results indicated that the concentrations of toxic trace elements of Ba, Cr, Se, Sn, Hg, Sin and Zn may represent pollutions problem

  10. Construction of prototype of on-line analyzer detection system for coal on belt conveyor using neutron activation technique

    International Nuclear Information System (INIS)

    Rony Djokorayono; Agus Cahyono; MP Indarzah; SG Usep; Sukandar

    2015-01-01

    The use of on-line neutron activation technique for coal analysis is proposed as an alternative method for analysis based on sampling technique. Compared to this conventional technique, the on-line neutron activation technique has much shorter time of analysis and more accurate results. The construction of detection system prototype for the on-line analyzer is described in this paper. This on-line analyzer consists of detection system, data acquisition system, and computer console. This detection system comprises several modules, i.e. NaI(Tl) scintillation detector completed with a photomultiplier tube (PMT), pre-amplifier, single channel analyzer (SCA), and analog signal transmitter and pulse counter processor. The construction processes of these four modules include the development of configuration block, lay out, and selection of electronic components. The modules have been integrated and tested. This detection system was tested using radioactive element Zn-65 having energy of 1115.5 keV and activity of 1 μCi. The test results show that the prototype of the on-line analyzer detection system has functioned as expected. (author)

  11. Feasibility of calibration of liquid sodium flowmeters by neutron activation techniques

    International Nuclear Information System (INIS)

    Kehler, P.

    1976-07-01

    Velocities of fluids in pipes can be measured by injecting radioactive tracers into the fluid and recording the activity downstream of the injection point. One convenient method of injecting radioactive tracers is by neutron activation of the fluid itself. The present report describes a FORTRAN program that can be used for the prediction of the counting rates of fluid flow tests performed with a pulsed neutron source and a scintillation detector. The program models the flow profile and the mixing of the fluid, the attenuation of neutrons and gamma rays in the fluid, and the geometric arrangement of the source and the detector. Using this program, an experiment for the measurement of the secondary sodium flow of the EBR-II was optimized. A pulsed D,T neutron source and a 5 in. x 5 in. NaI detector will be used in the EBR-II test. Under optimized conditions, the expected accuracy of the flow measurement is about 2 percent

  12. A comparison in the reconstruction of neutron spectrums using classical iterative techniques

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Vega C, H. R.; Gallego, E.

    2009-10-01

    One of the key drawbacks to the use of BUNKI code is that the process begins the reconstruction of the spectrum based on a priori knowledge as close as possible to the solution that is sought. The user has to specify the initial spectrum or do it through a subroutine called MAXIET to calculate a Maxwellian and a 1/E spectrum as initial spectrum. Because the application of iterative procedures by to resolve the reconstruction of neutron spectrum needs an initial spectrum, it is necessary to have new proposals for the election of the same. Based on the experience gained with a widely used method of reconstruction, called BUNKI, has developed a new computational tools for neutron spectrometry and dosimetry, which was first introduced, which operates by means of an iterative algorithm for the reconstruction of neutron spectra. The main feature of this tool is that unlike the existing iterative codes, the choice of the initial spectrum is performed automatically by the program, through a neutron spectra catalog. To develop the code, the algorithm was selected as the routine iterative SPUNIT be used in computing tool and response matrix UTA4 for 31 energy groups. (author)

  13. Implementation of variance-reduction techniques for Monte Carlo nuclear logging calculations with neutron sources

    NARCIS (Netherlands)

    Maucec, M

    2005-01-01

    Monte Carlo simulations for nuclear logging applications are considered to be highly demanding transport problems. In this paper, the implementation of weight-window variance reduction schemes in a 'manual' fashion to improve the efficiency of calculations for a neutron logging tool is presented.

  14. A new technique for the calibration of neutron probes by volumetric method

    International Nuclear Information System (INIS)

    Encarnacao, F.A.F. da.

    1988-01-01

    Laboratory and field studies were performed for the determination of a calibration curve of a neutron probe in three different kinds of soils: Red Yellow PODZOLIC, LITOLIC and ALLUVIAL, in the last one laboratory studies were done to determine local humidity on the calibration curve parameters. (A.C.A.S.) [pt

  15. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    2010-07-01

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  16. A novel Fe–Cr–Nb matrix composite containing the TiB_2 neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

    International Nuclear Information System (INIS)

    Litwa, Przemysław; Perkowski, Krzysztof; Zasada, Dariusz; Kobus, Izabela; Konopka, Gustaw; Czujko, Tomasz; Varin, Robert A.

    2016-01-01

    The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB_2 and a small quantity of Y_2O_3 ceramic to synthesize a novel Fe-based alloy-ceramic powder composite that could be processed by hot isostatic pressing (HIP) for a perceived potential application as a neutron absorber in nuclear reactors. After ball milling for the 30–80 h duration relatively uniform powders with micrometric sizes were produced. With increasing milling time a fraction of TiB_2 particles became covered with the much softer Fe-based alloy which resulted in the formation of a characteristic “core-mantel” structure. For the final HIP-ing process the mechanically alloyed powders were initially uniaxially pressed into rod-shaped compacts and then cold isostatically pressed (CIP-ed). Subsequently, the rod-shaped compacts were placed in the Ti-tubing and subjected to hot isostatic pressing (HIP) at 1150 °C/200 MPa pressure. The HIP-ing process resulted in the formation of the near-Ti and intermediate diffusional layers in the microstructure of HIP-ed samples which formed in accord with the Fe-Ti binary phase diagram. Those layers contain the phases such as α-Ti (HCP), the FeTi intermetallic and their hypo-eutectoid mixtures. In addition, needle-like particles were formed in both layers in accord with the Ti-B binary phase diagram. Nanohardness testing, using a Berkovich type diamond tip, shows that the nanohardness in the intermediate layer areas, corresponding to the composition of the hypo-eutectoid mixture of Ti-FeTi, equals 980.0 (±27.1) HV and correspondingly 1176.9 (±47.6) HV for the FeTi phase. The nanohardness in the sample's center in the areas with the fine mixture of Fe-based alloy and small TiB_2 particles equals 1048.3 (±201.8) HV. The average microhardness of samples HIP-ed from powders milled for 30 and 80 h is 588 HV and 733 HV, respectively. - Highlights: • A Fe–Cr–Nb-based composite with TiB_2 neutron absorbing ceramic was mechanically

  17. Use of the neutron diffraction - H/D exchange technique to determine the conformational dynamics of trypsin

    International Nuclear Information System (INIS)

    Kossiakoff, A.A.

    1982-01-01

    Reported here are studies analyzing the extent and nature of the inherent conformational fluctuations in trypsin by neutron diffraction - hydrogen exchange techniques. The pattern of exchange investigates systematic relationships between exchangeable sites and the structural and chemical properties of the molecule. Our findings that pH 7, 20 0 and 1 year of soaking all sites of trypsin are fully exchanged except those which are especially well protected by the structure. Essentially all the sites in which the peptide hydrogens are bonded directly to water molecules - either in the bulk solvent regions or in interior clusters - are fully exchanged. 41 references, 10 figures

  18. A rapid NiS bead technique for measurements of picogram concentrations of platinum group elements (PGEs) following neutron activation

    International Nuclear Information System (INIS)

    Das, A.; Shukla, A.D.

    1999-01-01

    To measure picogram amounts of PGEs in terrestrial and extraterrestrial samples we have modified the NiS fire assay technique in conjunction with neutron activation analysis. Os, Ir and Ru are almost quantitatively concentrated in the NiS bead. The method should be applicable to other elements (Pt, Pd, and Rh) but these could not be analyzed because of the short half life of their daughter isotopes. The results also show that the chalcophhile elements like Ag also can be quantitatively estimated using this method. (author)

  19. Determination of silver, gold, zinc and copper in mineral samples by various techniques of instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Rodriguez R, N. I.; Rios M, C.; Pinedo V, J. L.; Yoho, M.; Landsberger, S.

    2015-09-01

    Using the method of instrumental neutron activation analysis, mineral exploration samples were analyzed in order to determine the concentrations of silver, gold, zinc and copper; these minerals being the main products of benefit of Tizapa and Cozamin mines. Samples were subjected to various techniques, where the type of radiation and counting methods were chosen based on the specific isotopic characteristics of each element. For calibration and determination of concentrations the comparator method was used, certified standards were subjected to the same conditions of irradiation and measurement that the prospecting samples. The irradiations were performed at the research reactor TRIGA Mark II of the University of Texas at Austin. The silver concentrations were determined by Cyclical Epithermal Neutron Activation Analysis. This method in combination with the transfer pneumatic system allowed a good analytical precision and accuracy in prospecting for silver, from photo peak measurement 657.7 keV of short half-life radionuclide 110 Ag. For the determination of gold and zinc, Epithermal Neutron Activation Analysis was used, the photo peaks analyzed corresponded to the energies 411.8 keV of radionuclide 199 Au and 438.6 keV of metastable radionuclide 69m Zn. On the other hand, copper quantification was based on the photo peak analysis of 1039.2 keV produced by the short half-life radionuclide 66 Cu, by Thermal Neutron Activation Analysis. The photo peaks measurement corresponding to gold, zinc and copper was performed using a Compton suppression system, which allowed an improvement in the signal to noise relationship, so that better detection limits and low uncertainties associated with the results were obtained. Comparing elemental concentrations the highest values in silver, zinc and copper was for samples of mine Tizapa. Regarding gold values were found in the same range for both mines. To evaluate the precision and accuracy of the methods used, various geological

  20. Retinue of the beans roots growth by using neutron radiography technique

    International Nuclear Information System (INIS)

    Jesus, Selma Parente de; Crispim, Verginia Reis

    2002-01-01

    Agricultural practices frequently cause the development of a soil compacted layer below the surface. These compacted layers restrict the root penetration into deeper layers of soil, in search for water. It is proposed to monitor, using Non Destructive Test, the roots growth due to the planting of standard seeds in different agricultural soils, in function of their compactness and humidity. It will be used the neutrons beams derived from an irradiation channel called J-9 of the Reactor Argonauta (IEN/CNEN), so that the neutron radiographic images of the soil-plant system can be obtained. Each root can be evaluated for its ability to penetrate into compacted soil layers; this fact would mean an optimization of agricultural harvests. (author)

  1. Study of heterogeneous multiplying and non-multiplying media by the neutron pulsed source technique; Etude des milieux heterogenes multiplicateurs et non-multiplicateurs par la technique de la source pulsee de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Deniz, V [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-06-01

    The pulsed neutron technique consists essentially in sending in the medium to be studied a short neutron pulse and in determining the asymptotic decay constant of the generated population. The variation of the decay constant as a function of the size of the medium allows the medium characteristics to be defined. This technique has been largely developed these last years and has been applied as well to moderator as to multiplying media, in most cases homogeneous ones. We considered of interest of apply this technique to lattices, to see if useful informations could be collected for lattice calculations. We present here a general theoretical study of the problem, and results and interpretation of a series of experiments made on graphite lattices. There is a good agreement for non-multiplying media. In the case of multiplying media, it is shown that the age value used until now in graphite lattices calculations is over-estimated by about 10 per cent. [French] La technique de la pulsation neutronique consiste essentiellement a envoyer dans le milieu a etudier une courte bouffee de neutrons et a determiner la constante de decroissance asymptotique de la population engendree. La variation de cette constante de decroissance en fonction des dimensions du milieu permet de determiner ses caracteristiques. Cette technique a connu ces dernieres annees un grand essor et a ete appliquee a des moderateurs et des milieux multiplicateurs. Il s'agissait dans la plupart des cas de milieux homogenes. Il nous a semble interessant de l'utiliser dans le cas des reseaux, afin de voir si ces experiences peuvent fournir des renseignements utiles aux calculs. Nous presentons ici une etude theorique generale du probleme, ainsi que les resultats et l'interpretation d'une serie d'experiences faites sur des reseaux a graphite. L'accord est bon dans le cas des reseaux non-multiplicateurs. Dans le cas des reseaux multiplicateurs, on montre que la valeur de l'age utilisee jusqu'ici dans les calculs

  2. An empirical fast-neutron technique for detection of explosive-like materials

    International Nuclear Information System (INIS)

    Hussein, E.M.A.; Lord, P.M.; Bot, D.L.

    1990-01-01

    A method for detecting explosives in airport baggage using fast-neutron scattering and transmission measurements is presented. Ammonium nitrate (a commercial fertilizer) is used in the laboratory to simulate an explosive-like substance. The measurements are combined in Cartesian maps of normalized pairs of measurements. The existence of fertilizer manifests itself in these maps within a distinct region which is not significantly altered by the presence of surrounding materials. Monte Carlo simulations further confirm this phenomenon. (orig.)

  3. NCT-ART - a neutron computer tomography code based on the algebraic reconstruction technique

    International Nuclear Information System (INIS)

    Krueger, A.

    1988-01-01

    A computer code is presented, which calculates two-dimensional cuts of material assemblies from a number of neutron radiographic projections. Mathematically, the reconstruction is performed by an iterative solution of a system of linear equations. If the system is fully determined, clear pictures are obtained. Even for an underdetermined system (low number of projections) reasonable pictures are reconstructed, but then picture artefacts and convergence problems occur increasingly. (orig.) With 37 figs [de

  4. Determination of hydrogen content of Jatropha biodiesel oil using neutron reflection technique

    International Nuclear Information System (INIS)

    Okunade, I. O.; Jonah, S. A.; Omede, M.

    2014-01-01

    Biofuel is an environmental-friendly alternative to fossil fuel and holds immense potential for the future energy needs of the country. Non-edible jatropha biodiesel oil has been identified as one of the suitable bio-fuel options. In this work, experimental measurements were performed to determine the total hydrogen content of jatropha oil and jatropha-synthetic diesel oil mixture. The work was carried out using a neutron reflection facility at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria. Jatropha oil and jatropha-synthetic diesel mixture were subjected to experimental measurements for the purpose of determining bulk hydrogen content using neutron reflection facility that had been previously calibrated using various hydrocarbon materials of known hydrogen content. The hydrogen content of the sample were subsequently determined using their measured neutron reflection coefficient values and calibration data. In general, results obtained indicate high hydrogen content range of 10.68-12.16wt% for jatropha oil and the various jatropha-synthetic diesel mixtures. The implication of this is that jatropha oil or jatrophal-synthetic diesel mixture can be used as alternative fuel to mitigate high carbon monoxide emission.

  5. Application of the neutron activation analysis technique in trace elements analysis

    International Nuclear Information System (INIS)

    Khamis, I.; Sarheel, A.; Al-Somel, N.

    2006-12-01

    The main objective in this study is the implementation k 0 -standardization method (single comparator method) using gold comparator as a routine method in neutron activation analysis laboratory in Engineering Nuclear Department. Cadmium ratio Rcd; Cd-ratio = [A s p/(A s p) C d] and the nuclear reactor constants (f=φ t h/φ e pi subcadimum thermal- to- epithermal neutron flux ratio and α with describing the φ e (E)∼ 1/E 1+α neutron flux distribution) were determined in the inner and outer irradiation sites at MNS Reactor. K 0 -IAEA software, which provided by the Agency, has been installed and applied in our laboratory. Trace elements in many kinds of samples (biological, environmental, alloy ...etc) were determined using K 0 -IAEA software. The results of standard reference materials (SRM's) obtained in this work show a good agreement with the certified values, and we got these results with a good accuracy closer to results which we got from relative NAA method. (author)

  6. 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

  7. A newly developed technique of wireless remote controlled visual inspection system for neutron guides of cold neutron research facilities at HANARO

    International Nuclear Information System (INIS)

    Huh, Hyung; Cho, Yeong Garp; Kim, Jong In

    2012-01-01

    KAERI developed a neutron guide system for cold neutron research facilities at HANARO from 2003 to 2010. In 2008, the old plug shutter and instruments were removed, and a new plug and primary shutter were installed as the first cold neutron delivery system at HANARO. At the beginning of 2010, all the neutron guides and accessories had been successfully installed as well. The neutron guide system of HANARO consists of the in pile plug assembly with in pile guides, the primary shutter with in shutter guides, the neutron guides in the guide shielding room with secondary shutter, and the neutron guides in the neutron guide hall. Three kinds of glass materials were selected with optimum lengths by considering their lifetime, shielding, maintainability and cost as well. Radiation damage of the guides can occur on the coating and glass by neutron capturing in the glass. It is a big challenge to inspect a guide failure because of the difficult surrounding environment, such as high level radiation, limited working space, and massive hard work for removing and reinstalling the shielding blocks as shown in Fig 1. Therefore, KAERI has developed a wireless remote controlled visual inspection system for neutron guides using an infrared light camera mounted on the vehicle moving in the guide

  8. Comparison of neutron activation analysis techniques for the determination of uranium concentrations in geological and environmental materials

    International Nuclear Information System (INIS)

    Landsberger, S.; Kapsimalis, R.

    2013-01-01

    We have described the determination of uranium in environmental, geological, and agricultural specimens by three different non-destructive nuclear methods. The effectiveness, as defined as the lower limits of detection in this work, of quantifying trace levels of bulk uranium in geological samples was evaluated for several common NAA techniques. These techniques include short-lived and medium-lived neutron activation analysis using thermal and epithermal neutrons; these results were compared with an assessment of Compton suppressed gamma-ray counting. A careful evaluation of three major (n,γ) reactions with chlorine, manganese and sodium that could impede determining low levels of uranium due to high Compton continuums was done. The evaluation of Compton suppressed passive gamma counting revealed that uranium concentrations below 50 mg kg −1 were not adequate to achieve good counting statistics using the 234m Pa the second daughter product of 238 U. -- Highlights: ► Determination of uranium concentrations in geological, environmental, and agricultural specimens. ► Use of several NAA and passive counting methods. ► Identified several key interferences. ► Use of Compton suppression to minimize effects of interferences

  9. Applications of the associated-particle neutron-time-of-flight interrogation technique - From sheep to unexploded ordnance

    Energy Technology Data Exchange (ETDEWEB)

    Mitra, S. [Environmental Sciences Department, Brookhaven National Laboratory, Upton, NY 11973 and Nuclear Forensics R and D, Sandia National Laboratories, Albuquerque, NM 87185-0968 (United States)

    2013-04-19

    The associated-particle technique (APT) will be presented for some diverse applications that include on the one hand, analyzing the body composition of live sheep and on the other, identifying the fillers of unexploded ordnance (UXO). What began with proof-of-concept studies using a large laboratory based 14 MeV neutron generator of the 'associated-particle' type, soon became possible for the first time to measure total body protein, fat and water simultaneously in live sheep using a compact field deployable associated-particle sealed-tube neutron generator (APSTNG). This non-invasive technique offered the animal physiologist a tool to monitor the growth of an animal in response to new genetic, nutritional and pharmacologic methods for livestock improvement. While measurement of carbon (C), nitrogen (N) and oxygen (O) determined protein, fat and water because of the fixed stoichiometric proportions of these elements in these body components, the unique C/N and C/O ratios of high explosives revealed their identity in UXO. The algorithm that was developed and implemented to extract C, N and O counts from an APT generated gamma-ray spectrum will be presented together with the UXO investigations that involved preliminary proofof-concept studies and modeling with Monte Carlo produced synthetic spectra of 57-155 mm projectiles.

  10. Neutron diffusion in spheroidal, bispherical, and toroidal systems

    International Nuclear Information System (INIS)

    Williams, M.M.R.

    1986-01-01

    The neutron flux has been studied around absorbing bodies of spheroidal, bispherical, and toroidal shapes in an infinite nonabsorbing medium. Exact solutions have been obtained by using effective boundary conditions at the surfaces of the absorbing bodies. The problems considered are as follows: 1. Neutron flux and current distributions around prolate and oblate spheroids. It is shown that an equivalent sphere approximation can lead to accurate values for the rate of absorption. 2. Neutron flux and current in a bispherical system of unequal spheres. Three separate situations arise here: (a) two absorbing spheres, (b) two spherical sources, and (c) one spherical source and one absorbing sphere. It is shown how the absorption rate in the two spheres depends on their separation. 3. Neutron flux and current in a toroidal system: (a) an absorbing toroid and (b) a toroidal source. The latter case simulates the flux distribution from a thermonuclear reactor vessel. Finally, a brief description of how these techniques can be extended to multiregion problems is given

  11. High resolution neutron tomography applied to tooth fillings on real teeth by use of neutron lens

    International Nuclear Information System (INIS)

    Masschaele, B.; Cauwels, P.; Mondelaers, W.; Baechler, S.; Jolie, J.; Materna, T.

    2000-01-01

    Today tomography is a well known technique for nondestructive analysis of samples. By taking several X-ray pictures from an object, it is possible to make a 3D reconstruction. The same thing can be done with neutrons. Since very recent it is possible to produce a high-flux neutron beam. By looking at the attenuation of the neutron beam in the sample from different angles, it is possible to make a neutron tomography. The properties of neutrons are so much different from X-rays that a new era in tomography has started. Where X-rays have a hard time penetrating samples containing heavy elements (Pb, Bi, U, Hg, Au), neutrons just seem to walk through. But when the neutrons encounter samples containing light compounds like water, oil, paper, B, Li,... they are easily absorbed. This makes the use of neutrons for imaging complementary to the well known X-ray imaging. The most used tooth filling material nowadays is amalgam. Amalgam is a mixture of different metals, like silver, tin, copper, mercury. Mercury is dangerous for the human body when it enters the blood stream. These fillings are very dense and X-rays have a very hard time penetrating it. Neutrons are the ideal probe for investigation of these high density regions. The result of the tomography reveals information on the long term stability of amalgam fillings and could help the still ongoing debate on the safety of the fillings. (author)

  12. Application of sensitivity analysis to a simplified coupled neutronic thermal-hydraulics transient in a fast reactor using Adjoint techniques

    International Nuclear Information System (INIS)

    Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.

    2011-01-01

    In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)

  13. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Tobin, Stephen J.

    2011-01-01

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  14. Neutron and gamma-ray transmission technique for the on-line determination of moisture in coal and coke

    International Nuclear Information System (INIS)

    Sowerby, B.D.; Millen, M.J.; Rafter, P.T.

    1985-01-01

    A fast neutron and gamma-ray transmission technique is being developed for the on-line analysis of moisture in coal and coke. The technique utilises 252 Cf and 137 Cs sources and 3 He and NaI(Tl) detectors. Laboratory measurements on single coal samples have shown that moisture can be determined to better than 1 wt% over the range 0 to 16 wt% moisture and 5 to 17cm thickness. Reduced errors were obtained for restricted ranges of moisture and thickness. Preliminary measurements on coke of thickness 30 to 50cm have shown that moisture can be determined to within 0.26 wt% over the range 1 to 16 wt% moisture

  15. Advanced neutron and X-ray techniques for insights into the microstructure of EB-PVD thermal barrier coatings

    Energy Technology Data Exchange (ETDEWEB)

    Kulkarni, Anand [State University of New York, Stony Brook, NY 11794 (United States); Goland, Allen [State University of New York, Stony Brook, NY 11794 (United States); Herman, Herbert [State University of New York, Stony Brook, NY 11794 (United States)]. E-mail: hherman@ms.cc.sunysb.edu; Allen, Andrew J. [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Dobbins, Tabbetha [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); DeCarlo, Francesco [Argonne National Laboratory, Argonne, IL 60439 (United States); Ilavsky, Jan [Argonne National Laboratory, Argonne, IL 60439 (United States); Long, Gabrielle G. [Argonne National Laboratory, Argonne, IL 60439 (United States); Fang, Stacy [Chromalloy Gas Turbine Corporation, Orangeburg, NY 10962 (United States); Lawton, Paul [Chromalloy Gas Turbine Corporation, Orangeburg, NY 10962 (United States)

    2006-06-25

    The ongoing quest to increase gas turbine efficiency and performance (increased thrust) provides a driving force for materials development. While improved engine design and usage of novel materials provide solutions for increased engine operating temperatures, and hence fuel efficiency, reliability issues remain. Thermal barrier coatings (TBCs), deposited onto turbine components using the electron-beam physical vapor deposition (EB-PVD) process, exhibit unique pore architectures capable of bridging the technological gap between insulation/life extension and prime reliance. This article explores the potential of advanced X-ray and neutron techniques for comprehension of an EB-PVD TBC coating microstructure. While conventional microscopy reveals a hierarchy of voids, complementary advanced techniques allow quantification of these voids in terms of component porosities, anisotropy, size and gradient through the coating thickness. In addition, the derived microstructural parameters obtained both further knowledge of the nature and architecture of the porosity, and help establish its influence on the resultant thermal and mechanical properties.

  16. Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

    International Nuclear Information System (INIS)

    Freeman, Corey R.; Geist, William H.

    2010-01-01

    Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF 6 spins at high velocities in centrifuges to separate the molecules containing 238 U from those containing the lighter 235 U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF 6 gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

  17. Crystallite Size and Microstrain Measurement of Cathode Material after Mechanical Milling using Neutron Diffraction Technique

    Directory of Open Access Journals (Sweden)

    A. Fajar

    2010-12-01

    Full Text Available The measurements of neutron diffraction patterns of commercially product and 10 hour mechanically milled cathode material lithium cobaltites (LiCoO2 have been performed. Rietveld analysis using FullProf does not show the change of crystal structure due to milling process, but the diffraction pattern has a lower intensity and the diffraction-line was broadening. The results of line-broadening study on milled sample using Rietveld methods detected that the microstrain was occurred in the sample with value 0.127080(35 % and the average crystallite size was 424.78(20 Å.

  18. Studies on elemental analysis of Chinese traditional herbs by neutron activation technique and their mutagenic effect

    International Nuclear Information System (INIS)

    Hamzah, A.; Beh, C.W.; Abugassa, I.; Sarmani, S.B.; Liow, J.Y.

    2004-01-01

    Chinese herbs are accepted as an alternative medicine for specific treatment of illness. It is important to know the contents of these herbs that might cause gene mutation. Ten most popular herbs used in Malaysia were analyzed by instrumental neutron activation analysis. A total of 16 trace and major elements were determined and the concentration of elements varied depending on the origin of the herb. Toxic elements found in the samples were below the levels prescribed by health regulations. The mutagenicity test showed that there was no toxic effect due to the heavy metals present in the herbs. (author)

  19. Experimental and mathematical simulation techniques for determining an in-situ response testing method for neutron sensors used in reactor power plant protection systems

    International Nuclear Information System (INIS)

    Behbahani, A.

    1983-01-01

    An analytical neutron sensor response model and methods for transient response measurements of neutron sensors (compensated ionization chamber), including possible in-situ techniques have been developed and evaluated to meet the provisions of Draft Standard ISA Sd67.06, IEEE 338-1977, and NRC Guide 1.118. One in-situ method requires the perturbation of the high voltage detector (sensor) power supply and measurement of the sensor response. The response to a perturbation of the power supply is related to the response of the sensor to a transient change in neutron flux. Random signal analysis is another in-situ technique to monitor the neutron sensor response. In this method the power spectrum of the inherent fluctuations from the neutron sensor output (current in CIC) are measured and evaluated. Transient response techniques (including in-situ methods) are experimentally and analytically evaluated to identify the mechanisms which may cause degradation in the response of the neutron sensors. The objective of the experimental evaluation was to correlate the measured response time using transient radiation flux changes and power supply perturbation. The objective of the analytical model of the different sensor response was to predict response time and degradation mechanisms

  20. 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.