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Sample records for absorber neutronics performance

  1. 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.)

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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)

  12. 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.)

  13. 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.)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. 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.

  2. 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)

  3. 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

  4. 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)

  5. 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.)

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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)

  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. 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.

  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. 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.)

  3. 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)

  4. 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

  5. A proposal on evaluation method of neutron absorption performance to substitute conventional neutron attenuation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Je Hyun; Shim, Chang Ho [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of); Kim, Sung Hyun [Nuclear Fuel Cycle Waste Treatment Research Division, Research Reactor Institute, Kyoto University, Osaka (Japan); Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo [Ionizing Radiation Center, Nuclear Fuel Cycle Waste Treatment Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho [Ionizing Radiation Center, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

    2016-12-15

    For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

  6. A proposal on evaluation method of neutron absorption performance to substitute conventional neutron attenuation test

    International Nuclear Information System (INIS)

    Kim, Je Hyun; Shim, Chang Ho; Kim, Sung Hyun; Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo; Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho

    2016-01-01

    For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers

  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. 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)

  10. 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

  11. 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

  12. 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

  13. Neutronic performance issues for the Spallation Neutron Source moderators

    International Nuclear Information System (INIS)

    Iverson, E.B.; Murphy, B.D.

    2001-01-01

    We continue to develop the neutronic models of the Spallation Neutron Source target station and moderators in order to better predict the neutronic performance of the system as a whole and in order to better optimize that performance. While we are not able to say that every model change leads to more intense neutron beams being predicted, we do feel that such changes are advantageous in either performance or in the accuracy of the prediction of performance. We have computationally and experimentally studied the neutronics of hydrogen-water composite moderators such as are proposed for the SNS Project. In performing these studies, we find that the composite moderator, at least in the configuration we have examined, does not provide performance characteristics desirable for the instruments proposed and being designed for this neutron scattering facility. The pulse width as a function of energy is significantly broader than for other moderators, limiting attainable resolution-bandwidth combinations. Furthermore, there is reason to expect that higher-energy (0.1-1 eV) applications will be significantly impacted by bimodal pulse shapes requiring enormous effort to parameterize. As a result of these studies, we have changed the SNS design, and will not use a composite moderator at this time. We have analyzed the depletion of a gadolinium poison plate in a hydrogen moderator at the Spallation Neutron Source, and found that conventional poison thicknesses will be completely unable to last the desired component lifetime of three operational years. A poison plate 300-600 μm thick will survive for the required length of time, but will somewhat degrade the intensity (by as much as 15% depending on neutron energy) and the consistency of the neutron source performance. Our results should scale fairly easily to other moderators on this or any other spallation source. While depletion will be important for all highly-absorbing materials in high-flux regions, we feel it likely that

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. 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)

  3. 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)

  4. Performance of an absorbing concentrating solar collectors

    International Nuclear Information System (INIS)

    Imadojemu, H.

    1990-01-01

    This paper reports on a comparison of the efficiency of an absorbing fluid parabolic trough concentrating solar collector and a traditional concentrating collector that was made. In the absorbing fluid collector, black liquid flows through a glass tube absorber while the same black liquid flows through a selective black coated copper tube absorber while the same black fluid flows through a selective black coated copper tube absorber in the traditional collector. After a careful study of the properties of available black liquids, a mixture of water and black ink was chosen as the black absorbing medium or transfer fluid. In the black liquid glass collector there is a slightly improved efficiency based on beam radiation as a result of the direct absorption process and an increase in the effective transmittance absorptance. At worst the efficiency of this collector equals that of the traditional concentrating collector when the efficiency is based on total radiation. The collector's reflecting surfaces were made of aluminum sheet, parabolic line focus and with cylindrical receivers. The ease of manufacture and reduced cost per unit energy collected, in addition to the clean and pollution free mode of energy conversion, makes it very attractive

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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)

  15. Performance of Closely Spaced Point Absorbers with Constrained Floater Motion

    DEFF Research Database (Denmark)

    De Backer, G.; Vantorre, M.; Beels, C.

    2009-01-01

    The performance of a wave energy converter array of twelve heaving point absorbers has been assessed numerically in a frequency domain model. Each point absorber is assumed to have its own linear power take-off. The impact of slamming, stroke and force restrictions on the power absorption...

  16. 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

  17. Optimization of sound absorbing performance for gradient multi-layer-assembled sintered fibrous absorbers

    Science.gov (United States)

    Zhang, Bo; Zhang, Weiyong; Zhu, Jian

    2012-04-01

    The transfer matrix method, based on plane wave theory, of multi-layer equivalent fluid is employed to evaluate the sound absorbing properties of two-layer-assembled and three-layer-assembled sintered fibrous sheets (generally regarded as a kind of compound absorber or structures). Two objective functions which are more suitable for the optimization of sound absorption properties of multi-layer absorbers within the wider frequency ranges are developed and the optimized results of using two objective functions are also compared with each other. It is found that using the two objective functions, especially the second one, may be more helpful to exert the sound absorbing properties of absorbers at lower frequencies to the best of their abilities. Then the calculation and optimization of sound absorption properties of multi-layer-assembled structures are performed by developing a simulated annealing genetic arithmetic program and using above-mentioned objective functions. Finally, based on the optimization in this work the thoughts of the gradient design over the acoustic parameters- the porosity, the tortuosity, the viscous and thermal characteristic lengths and the thickness of each samples- of porous metals are put forth and thereby some useful design criteria upon the acoustic parameters of each layer of porous fibrous metals are given while applying the multi-layer-assembled compound absorbers in noise control engineering.

  18. 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

  19. Performance evaluation of CFRP-rubber shock absorbers

    Science.gov (United States)

    Lamanna, Giuseppe; Sepe, Raffaele

    2014-05-01

    In the present work a numerical investigation on the energy absorbing capability of dedicated structural components made of a carbon fiber reinforced polymer and an emulsion polymerised styrene butadiene rubber is reported. The shock absorbers are devices designed to absorb large amounts of energy by sacrificing their own structural integrity. Their aim is to cushion the effects of an impact phenomenon with the intent to preserve other structures from global failure or local damaging. Another important role of shock absorbers is reducing the peak of the acceleration showed during an impact phenomenon. This effect is of considerable interest in the case of vehicles to preserve passengers' safety. Static and dynamic numerical results are compared with experimental ones in terms of mean crushing forces, energy and peak crushing. The global performance of the absorbers has been evaluated by referencing to a proposed quality index.

  20. Performance evaluation of CFRP-rubber shock absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Lamanna, Giuseppe, E-mail: giuseppe.lamanna@unina2.it; Sepe, Raffaele, E-mail: giuseppe.lamanna@unina2.it [Department of Industrial and Information Engineering, Second University of Naples, via Roma, 29 - 81031 Aversa (Italy)

    2014-05-15

    In the present work a numerical investigation on the energy absorbing capability of dedicated structural components made of a carbon fiber reinforced polymer and an emulsion polymerised styrene butadiene rubber is reported. The shock absorbers are devices designed to absorb large amounts of energy by sacrificing their own structural integrity. Their aim is to cushion the effects of an impact phenomenon with the intent to preserve other structures from global failure or local damaging. Another important role of shock absorbers is reducing the peak of the acceleration showed during an impact phenomenon. This effect is of considerable interest in the case of vehicles to preserve passengers’ safety. Static and dynamic numerical results are compared with experimental ones in terms of mean crushing forces, energy and peak crushing. The global performance of the absorbers has been evaluated by referencing to a proposed quality index.

  1. Corrosion behaviour of borated aluminium used as neutron absorber

    Energy Technology Data Exchange (ETDEWEB)

    Emmerich, R. [EaglePicher Technologies GmbH, Oehringen (Germany); Ensinger, W.; Enders, B. [Philipps-Univ. of Marburg, Dept. of Chemistry, Material Science Centre (Germany)

    2004-07-01

    The electrochemical behaviour of pure and borated aluminium was examined. Measurements were performed in two different electrolytes at 90 C containing different trace-amounts of chloride. For borated aluminium current transients, i.e. metastable depassivation events were found. It is suggested to attribute these transients to less stable passivation layers in comparison to pure aluminium.

  2. Enhancement of acoustical performance of hollow tube sound absorber

    International Nuclear Information System (INIS)

    Putra, Azma; Khair, Fazlin Abd; Nor, Mohd Jailani Mohd

    2016-01-01

    This paper presents acoustical performance of hollow structures utilizing the recycled lollipop sticks as acoustic absorbers. The hollow cross section of the structures is arranged facing the sound incidence. The effects of different length of the sticks and air gap on the acoustical performance are studied. The absorption coefficient was measured using impedance tube method. Here it is found that improvement on the sound absorption performance is achieved by introducing natural kapok fiber inserted into the void between the hollow structures. Results reveal that by inserting the kapok fibers, both the absorption bandwidth and the absorption coefficient increase. For test sample backed by a rigid surface, best performance of sound absorption is obtained for fibers inserted at the front and back sides of the absorber. And for the case of test sample with air gap, this is achieved for fibers introduced only at the back side of the absorber.

  3. Enhancement of acoustical performance of hollow tube sound absorber

    Energy Technology Data Exchange (ETDEWEB)

    Putra, Azma, E-mail: azma.putra@utem.edu.my; Khair, Fazlin Abd, E-mail: fazlinabdkhair@student.utem.edu.my; Nor, Mohd Jailani Mohd, E-mail: jai@utem.edu.my [Centre for Advanced Research on Energy, Universiti Teknikal Malaysia Melaka, Hang Tuah Jaya, Durian Tunggal Melaka 76100 Malaysia (Malaysia)

    2016-03-29

    This paper presents acoustical performance of hollow structures utilizing the recycled lollipop sticks as acoustic absorbers. The hollow cross section of the structures is arranged facing the sound incidence. The effects of different length of the sticks and air gap on the acoustical performance are studied. The absorption coefficient was measured using impedance tube method. Here it is found that improvement on the sound absorption performance is achieved by introducing natural kapok fiber inserted into the void between the hollow structures. Results reveal that by inserting the kapok fibers, both the absorption bandwidth and the absorption coefficient increase. For test sample backed by a rigid surface, best performance of sound absorption is obtained for fibers inserted at the front and back sides of the absorber. And for the case of test sample with air gap, this is achieved for fibers introduced only at the back side of the absorber.

  4. 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

  5. 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)

  6. 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

  7. 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

    , which greatly reduced the cumbersome administrative controls and limits. One of the engineering design concepts for the new racks consists of one with 20 horizontal storage positions. Table I summarizes the results for this storage rack for three cases. The first column ignores neutron absorbers in the structural materials, and 70 kg of plutonium cannot be stored in a cubicle. The second column allows for the presence of neutron absorbers in structural and can materials; however, only ten positions can contain 4.4 kg of plutonium as a compound, and ten positions can hold containers with 2.5 kg of plutonium. The third column considers the presence of the neutron absorber plate. There are no restrictions on the location of the 4.4-kg plutonium containers in the rack. The neutron absorber plate needs to have a concentration of 0.18 g/cm 2 of 10 B. The calculations performed with the MCNP4B code are all based on a single-cubicle model with reflecting boundary conditions to simulate an infinite planar array of cubicles. The reflected single-cubicle model was shown to be a conservative design configuration for the vault analysis. It provided calculated k eff 's with a delta k eff = 0.02 higher than a full vault model with 68 cubicles. With no absorbers considered, a k eff = 0.943 was calculated, which exceeded the target criticality limit of k eff = 0.942. Inclusion of absorbers in the structural components reduced the k eff to 0.926 for contingency conditions. With the neutron absorber plate present, a k eff of 0.885 was achieved. When all cans containing 2.5 kg of plutonium are replaced by cans loaded with 4.4 kg of plutonium, the k eff = 0.933 for 88 kg of plutonium in the cubicle. The presence of neutron absorber verification as required by ANS 8.21 (Ref. 3) can readily be accomplished by visual inspections in the cubicle geometry. The analysis presented in this paper demonstrates the importance of engineered safety features as manifested by the inclusion of neutron

  8. Parametric study on the performance of automotive MR shock absorbers

    Science.gov (United States)

    Gołdasz, J.; Dzierżek, S.

    2016-09-01

    The paper contains the results of a parametric study to explore the influence of various quantities on the performance range of semi-active automotive shock absorbers using the magnetorheological (MR) fluid under steady-state and transient excitations. The analysis was performed with simulated data and using a standard single-tube shock absorber configuration with a single-gap MR valve. Additionally, the impact of material variables and valves geometry was examined as the parameters were varied and its dynamic range studied.

  9. 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

  10. 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.

  11. Absorber performance of a water/lithium-bromide absorption chiller

    International Nuclear Information System (INIS)

    Xie Guozhen; Sheng Guogang; Bansal, Pradeep Kumar; Li, Guang

    2008-01-01

    An absorber is one of the most important components of a lithium-bromide absorption chiller (LBAC) as its absorbing characteristics directly influence the performance of the whole chiller. It has been indicated that the absorbing efficiency and cooling capacity could be improved by increasing the solution concentration. In this paper, based on the mechanism of falling film absorption on horizontal tubes, the theoretical models of falling film absorption on horizontal tubes have been established. A series of programs used for computing the theoretical mathematical models, including simulation of LBAC cycle and falling film absorption, have been programmed. The models have been validated reasonably by the experimental data. The results show that the cooling capacity of the LBAC varies in parabola shape of curve with the solution concentration from 52.5% to 58.5%, and that the best coefficient of performance (COP) occurs at concentration of 57%. The investigation proposes the absorbing process of sub-steady thermodynamic equilibrium for the duality solution under increase absorbing pressure

  12. 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}.

  13. 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

  14. Performance of a Multifunctional Space Evaporator-Absorber-Radiator (SEAR)

    Science.gov (United States)

    Izenson, Michael G.; Chen, Weibo; Phillips, Scott; Chepko, Ariane; Bue, Grant; Quinn, Gregory

    2014-01-01

    The Space Evaporator-Absorber-Radiator (SEAR) is a nonventing thermal control subsystem that combines a Space Water Membrane Evaporator (SWME) with a Lithium Chloride Absorber Radiator (LCAR). The LCAR is a heat pump radiator that absorbs water vapor produced in the SWME. Because of the very low water vapor pressure at equilibrium with lithium chloride solution, the LCAR can absorb water vapor at a temperature considerably higher than the SWME, enabling heat rejection sufficient for most EVA activities by thermal radiation from a relatively small area radiator. Prior SEAR prototypes used a flexible LCAR that was designed to be installed on the outer surface of a portable life support system (PLSS) backpack. This paper describes a SEAR subsystem that incorporates a very compact LCAR. The compact, multifunctional LCAR is built in the form of thin panels that can also serve as the PLSS structural shell. We designed and assembled a 2 ft² prototype LCAR based on this design and measured its performance in thermal vacuum tests when supplied with water vapor by a SWME. These tests validated our models for SEAR performance and showed that there is enough area available on the PLSS backpack shell to enable rejection of metabolic heat from the LCAR. We used results of these tests to assess future performance potential and suggest approaches for integrating the SEAR system with future space suits.

  15. 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

  16. 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.

  17. 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.

  18. 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).

  19. 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

  20. 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

  1. 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)

  2. 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)

  3. {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

  4. 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

  5. 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

  6. Load testing of neutron absorbent materials by using neutron radiographic pictures

    International Nuclear Information System (INIS)

    Bayon, G.; Laporte, A.

    1986-06-01

    By using standards whose characteristics are known, quantitative measurement of the neutronographic negatives density, makes possible the estimation of a neutrophagic constituent proportion in each point of a given material. This presentation describes the methodology, the automatic equipment developed at Saclay and the method performance characteristics applied to mass-produced components

  7. 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.

  8. The minireactor Mirene for neutron-radiography: performances and applications

    International Nuclear Information System (INIS)

    Houelle, M.; Gerberon, J.M.

    1981-05-01

    The MIRENE neutron radiograhy mini-reactor is described. The core contains only one kilogram of enriched uranium in solution form. It works by pulsed operation. The neutron bursts produced are collimated into two beams which pass through the concrete protection around the reactor block. The performance of the reactor and the results achieved since it went into service in 1977 are described. These concern various fields. In the nuclear field: examination of fast neutron reactor fissile pins, monitoring of neutron absorbing screens employed to guarantee the safety-criticality of the transport and storage of the nuclear fuel cycle, observation of irradiated oxide fuel pellets in order to determine the fuel state equation of the fast neutron system, examination of UO 2 and water mixtures for criticality experiments. In the industrial field, Mirene has a vast field of application. Two examples are given: monitoring of electric insulation sealing, visualization of the bonding of two high density metal parts. Finally an original application in agronomy has given very good results: this concerns the on-site follow-up of the root growth of maize plants [fr

  9. 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

  10. 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.

  11. 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)

  12. A neutron irradiator to perform nuclear activation

    International Nuclear Information System (INIS)

    Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

    2001-01-01

    The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

  13. Study of influence of transport performance of the neutron guide

    International Nuclear Information System (INIS)

    Li Xinxi; Wang Yan; Huang Chaoqiang; Chen Bo; Chen Liang

    2009-01-01

    For the sake of improving the performance of the neutron scattering instrument, usually we need use the neutron guide, it's very important to select the right type and optimizing of neutron guide. The papers calculate the focus neutron guide and the single channel neutron guide by numeric method. The results shows that the choice of neutron guide should consult the resolution requirement of neutron scattering instrument, and the length of the neutron guide should be optimized. The calculation results can be the theoretical reference for the design of neutron scattering instrument. (authors)

  14. 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

  15. 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...

  16. Materials performance experience at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Sommer, W.F. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    There is a growing, but not yet substantial, data base for materials performance at spallation neutron sources. Specially designed experiments using medium energy protons (650 MeV) have been conducted at the Proton Irradiation Experiment (PIREX) facility at the Swiss Nuclear Institute accelerator (SIN). Specially designed experiments using 760-800 MeV copper target have been completed at the Los Alamos Spallation Radiation Effects Facility (LASREF) at Los Alamos Meson Physics Facility (LAMPF). An extensive material testing program was initiated at LASREF in support of the German spallation neutron source (SNQ) project, before it terminated in 1985.

  17. Neutron performance analysis for ESS target proposal

    International Nuclear Information System (INIS)

    Magán, M.; Terrón, S.; Thomsen, K.; Sordo, F.; Perlado, J.M.; Bermejo, F.J.

    2012-01-01

    In the course of discussing different target types for their suitability in the European Spallation Source (ESS) one main focus was on neutronics' performance. Diverse concepts have been assessed baselining some preliminary engineering and geometrical details and including some optimization. With the restrictions and resulting uncertainty imposed by the lack of detailed designs optimizations at the time of compiling this paper, the conclusion drawn is basically that there is a little difference in the neutronic yield of the investigated targets. Other criteria like safety, environmental compatibility, reliability and cost will thus dominate the choice of an ESS target.

  18. 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.

  19. 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.)

  20. 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.

  1. 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.

  2. 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

  3. 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)

  4. 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.

  5. 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

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. Eurados trial performance test for neutron personal dosimetry

    DEFF Research Database (Denmark)

    Bordy, J.M.; Stadtmann, H.; Ambrosi, P.

    2001-01-01

    This paper reports on the results of a neutron trial performance test sponsored by the European Commission and organised by EURADOS. As anticipated, neutron dosimetry results were very dependent on the dosemeter type and the dose calculation algorithm. Fast neutron fields were generally well...

  12. 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.

  13. Performance of an elliptically tapered neutron guide

    International Nuclear Information System (INIS)

    Muehlbauer, Sebastian; Stadlbauer, Martin; Boeni, Peter; Schanzer, Christan; Stahn, Jochen; Filges, Uwe

    2006-01-01

    Supermirror coated neutron guides are used at all modern neutron sources for transporting neutrons over large distances. In order to reduce the transmission losses due to multiple internal reflection of neutrons, ballistic neutron guides with linear tapering have been proposed and realized. However, these systems suffer from an inhomogeneous illumination of the sample. Moreover, the flux decreases significantly with increasing distance from the exit of the neutron guide. We propose using elliptically tapered guides that provide a more homogeneous phase space at the sample position as well as a focusing at the sample. Moreover, the design of the guide system is simplified because ellipses are simply defined by their long and short axes. In order to prove the concept we have manufactured a doubly focusing guide and investigated its properties with neutrons. The experiments show that the predicted gains using the program package McStas are realized. We discuss several applications of elliptic guides in various fields of neutron physics

  14. Performance of self-powered neutron detectors in pressurized water reactors

    International Nuclear Information System (INIS)

    Warren, H.D.; Bozarch, D.P.

    1977-01-01

    A typical Babcock and Wilcox pressurized water reactor (PWR) contains 364 rhodium self-powered neutron detectors (SPNDs) and 52 background detectors. The detectors are inserted into the reactor core in 52 dry, multidetector assemblies. Each assembly contains seven SPNDs and one background detector. By mid-1977, eight B and W PWRs, each fitted with SPNDs, were in operation. Many of the SPNDs have operated successfully for more than four years. This paper describes the operational performance of the SPNDs and special tests conducted to improve that performance. Topics included are (1) insulation performance versus neutron dose to the SPND, (2) background signals in the leadwire region of the SPND, and (3) depletion of the SPND emitter versus absorbed neutron dose

  15. 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

  16. Exergy Based Performance Analysis of Double Flow Solar Air Heater with Corrugated Absorber

    OpenAIRE

    S. P. Sharma; Som Nath Saha

    2017-01-01

    This paper presents the performance, based on exergy analysis of double flow solar air heaters with corrugated and flat plate absorber. A mathematical model of double flow solar air heater based on energy balance equations has been presented and the results obtained have been compared with that of a conventional flat-plate solar air heater. The double flow corrugated absorber solar air heater performs thermally better than the flat plate double flow and conventional flat-plate solar air heate...

  17. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  18. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  19. Neutron shielding performance of water-extended polyester

    Energy Technology Data Exchange (ETDEWEB)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  20. 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)

  1. 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

  2. Performance Study of Solar Air Heater Having Absorber Plate with Half-Perforated Baffles

    OpenAIRE

    Maheshwari, B. K.; Karwa, Rajendra; Gharai, S. K.

    2011-01-01

    The paper presents a detailed mathematical model for performance prediction of a smooth duct solar air heater validated against the experimental results. Experimental study on a solar air heater having absorber plate with half-perforated baffles on the air flow side shows thermal efficiency enhancement of 28%–45% over that of the smooth duct solar air heater, which is attributed to the heat transfer enhancement (of the order of 180%–235%) due to the perforated baffles attached to the absorber...

  3. Performance of a thermal neutron radiographic system using imaging plates

    International Nuclear Information System (INIS)

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

    2009-01-01

    A performance evaluation of a neutron radiographic system equipped with a thermal neutron sensitive imaging plate has been undertaken. It includes the assessment of spatial resolution, linearity, dynamic range and the response to exposure time, as well as a comparison of these parameters with the equivalent ones for neutron radiography employing conventional films and a gadolinium foil as converter. The evaluation and comparison between the radiographic systems have been performed at the Instituto de Engenharia Nuclear - CNEN, using the Argonauta Reactor as source of thermal neutrons and a commercially available imaging plate reader. (author)

  4. Effect of Collector Aspect Ratio on the Thermal Performance of Wavy Finned Absorber Solar Air Heater

    OpenAIRE

    Abhishek Priyam; Prabha Chand

    2016-01-01

    A theoretical investigation on the effect of collector aspect ratio on the thermal performance of wavy finned absorber solar air heaters has been performed. For the constant collector area, the various performance parameters have been calculated for plane and wavy finned solar air heaters. It has been found that the performance of wavy finned solar air heater improved with the increase in the collector aspect ratio. The performance of wavy finned solar air heater has been found 30 percent hig...

  5. 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))

  6. Neutron metrology file NMF-90. An integrated database for performing neutron spectrum adjustment calculations

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1996-01-01

    The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs

  7. Albedo neutron dosimetry in Germany: regulations and performance

    International Nuclear Information System (INIS)

    Luszik-Bhadra, M.; Zimbal, A.; Busch, F.; Jordan, M.; Eichelberger, A.; Engelhardt, J.; Martini, E.; Figel, M.; Haninger, T.; Frasch, G.; Guenther, K.; Seifert, R.; Rimpler, A.

    2014-01-01

    Personal neutron dosimetry has been performed in Germany using albedo dosemeters for >20 y. This paper describes the main principles, the national standards, regulations and recommendations, the quality management and the overall performance, giving some examples. (authors)

  8. 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

  9. Performance of neutron polarimeter SMART-NPOL

    International Nuclear Information System (INIS)

    Noji, S.; Miki, K.; Yako, K.; Kawabata, T.; Kuboki, H.; Sakai, H.; Sekiguchi, K.; Suda, K.

    2007-01-01

    The neutron polarimeter SMART-NPOL has been constructed at the RIKEN Accelerator Research Facility for measuring polarization correlations of proton-neutron systems. The SMART-NPOL system consists of 12 parallel neutron counter planes of two dimensionally position-sensitive plastic scintillators with a size of 60x60x3.0cm 3 . Polarimetry measurements were made using the analyzing power of the H1(n-vector,n)H1 reaction occurring in the plastic scintillators. The effective analyzing power of SMART-NPOL was measured with polarized neutrons from the zero-degree Li6(d-vector,n-vector) reaction with an incident deuteron energy of 135MeV/A. The effective analyzing power thus obtained was 0.26±0.01 stat ±0.03 syst and the double scattering efficiency was 1.1x10 -3

  10. Neutronic performances of the MEGAPIE target

    Energy Technology Data Exchange (ETDEWEB)

    Panebianco, S.; Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee (DSM/DAPNIA/SPhN), 91- Gif sur Yvette (France); Beauvais, P.; Lotrus, P.; Molinie, F.; Toussaint, J.Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee (DSM/DAPNIA), 91- Gif sur Yvette (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie (DEN/DPC/SECR), 91 - Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs (DEN/DER/SPEX), 13 - Saint Paul lez Durance (France)

    2008-07-01

    The MEGAPIE project is a key experiment on the road to Accelerator Driven Systems and it provides the scientific community with unique data on the behavior of a liquid lead-bismuth spallation target under realistic and long term irradiation conditions. The neutronic of such target is of course of prime importance when considering its final destination as an intense neutron source. This is the motivation to characterize the inside neutron flux of the target in operation. A complex detector, made of 8 'micro' fission-chambers, has been built and installed in the core of the target, few tens of centimeters from the proton/Pb-Bi interaction zone. This detector is designed to measure the absolute neutron flux inside the target, to give its spatial distribution and to correlate its temporal variations with the beam intensity. Moreover, integral information on the neutron energy distribution as a function of the position along the beam axis could be extracted, giving integral constraints on the neutron production models implemented in transport codes such as MCNPX. (authors)

  11. A review on the performance of conventional and energy-absorbing rockbolts

    Directory of Open Access Journals (Sweden)

    Charlie C. Li

    2014-08-01

    Full Text Available This is a review paper on the performances of both conventional and energy-absorbing rockbolts manifested in laboratory tests. Characteristic parameters such as ultimate load, displacement and energy absorption are reported, in addition to load–displacement graphs for every type of rockbolt. Conventional rockbolts refer to mechanical rockbolts, fully-grouted rebars and frictional rockbolts. According to the test results, under static pull loading a mechanical rockbolt usually fails at the plate; a fully-grouted rebar bolt fails in the bolt shank at an ultimate load equal to the strength of the steel after a small amount of displacement; and a frictional rockbolt is subjected to large displacement at a low yield load. Under shear loading, all types of bolts fail in the shank. Energy-absorbing rockbolts are developed aiming to combat instability problems in burst-prone and squeezing rock conditions. They absorb deformation energy either through ploughing/slippage at predefined load levels or through stretching of the steel bolt. An energy-absorbing rockbolt can carry a high load and also accommodate significant rock displacement, and thus its energy-absorbing capacity is high. The test results show that the energy absorption of the energy-absorbing bolts is much larger than that of all conventional bolts. The dynamic load capacity is smaller than the static load capacity for the energy-absorbing bolts displacing based on ploughing/slippage while they are approximately the same for the D-Bolt that displaces based on steel stretching.

  12. Performance Evaluation of the New Fork-Absorbers of RSG-GAS Control Rod

    International Nuclear Information System (INIS)

    Slamet Wiranto; Purwadi; Arif Hidayat; Agus Sanjaya

    2012-01-01

    During the operation of RSG-GAS reactor, it has been replaced 8 fork-absorber by the new absorber from PT. Batan Teknologi. After almost 5 years under utilization it is important to be evaluated to determine the physical condition and its performance, which is still in good condition and functioning according to the requirements of its operations. The evaluation has been carried out by studying and analyzing the data of the fork-absorber utilization in the the reactor core. The fork absorber data consist of visual inspection, control rod drop time measurement and control rod reactivity and safety margin measurement for each operation cycle. Through the observation up to date with the operating cycle of 79, could be concluded that the fork-absorber condition is still good, and has ability, to support the operation until ± 660 MWD/cycle, which is characterized by obtaining the value of ρ-excess is sufficient for operation, with a large safety margin. (author)

  13. Performance Test of BF3 Neutron Detection System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Shin, Ho Cheol [KHNP-CRI, Daejeon (Korea, Republic of); Cho, Jin Bok; Oh, Sae Hyun; Ryou, Seok Jean [USERS, Daejeon (Korea, Republic of)

    2015-10-15

    The neutron detecting system of First-of-a-kind plant such an APR1400 at Shin Kori should have been verified in the condition of low operating temperature and pressure of the primary coolant system before receiving the operation license. Auxiliary Ex-core Neutron Flux Monitoring System (AENFMS) is supposed to be installed using BF3 neutron detector in Shin Kori plant. The performance test of AENFMS was conducted to measure neutron sensitivity, moderation ratio and count rate in the same condition with Ex-core Neutron Flux Monitoring System (ENFMS) of APR1400 to verify its detection characteristics in compliance with the functional requirement. Performance test has been conducted for AENFMS of APR1400 to verify BF3 neutron sensitivity, moderation ration of PE, expecting neutron signal count rate from AENFMS, possible extending cable length from detector to pre-amplifier. As a result of measurement, the neutron sensitivity of 34.246±0.168(95%CI)cps/nv, moderation ratio of 11.343±0.039(95%CI) and AENFMS expecting count rate related to ENFMS of 17.8 times are acceptable in compliance with functional requirement, respectively.

  14. Roof Integrated Solar Absorbers: The Measured Performance of ''Invisible'' Solar Collectors: Preprint

    International Nuclear Information System (INIS)

    Colon, C. J.; Merrigan, T.

    2001-01-01

    The Florida Solar Energy Center (FSEC), with the support of the National Renewable Energy Laboratory, has investigated the thermal performance of solar absorbers that are an integral, yet indistinguishable, part of a building's roof. The first roof-integrated solar absorber (RISA) system was retrofitted into FSEC's Flexible Roof Facility in Cocoa, Florida, in September 1998. This ''proof-of-concept'' system uses the asphalt shingle roof surface and the plywood decking under the shingles as an unglazed solar absorber. Data was gathered for a one-year period on the system performance. In Phase 2, two more RISA prototypes were constructed and submitted for testing. The first used the asphalt shingles on the roof surface with the tubing mounted on the underside of the plywood decking. The second prototype used metal roofing panels over a plywood substrate and placed the polymer tubing between the plywood decking and the metal roofing. This paper takes a first look at the thermal performance results for the ''invisible'' solar absorbers that use the actual roof surface of a building for solar heat collection

  15. Performance review: neutron hodoscope at TREAT

    International Nuclear Information System (INIS)

    Stanford, G.S.; DeVolpi, A.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stewart, R.R.

    1976-01-01

    The current fuel-motion-detection capabilities of the neutron hodoscopes at TREAT are outlined and discussed, including such topics as spatial and fuel-density resolution, dynamic range, and corrections for detector dead time and supralinearity. Capabilities and analytical techniques are illustrated with examples from several of the power-transient experiments that have been run in the TREAT reactor

  16. The neutronic performance of solid-target alternatives for SINQ

    International Nuclear Information System (INIS)

    Atchison, F.

    1991-01-01

    The results from calculations of the neutronic performance of three possible 'solid' targets and that of the current version of the liquid Pb-Bi target are presented. Two are 'conventional' transverse cooled plate structures, one using tantalum, the other tungsten. The third is a Pb-shot based pebble-bed design. Some general results on the effect of neutron absorption on the performance of the Pebble-bed target are given. (author)

  17. 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)

  18. Performance evaluation and parameter sensitivity of energy-harvesting shock absorbers on different vehicles

    Science.gov (United States)

    Guo, Sijing; Liu, Yilun; Xu, Lin; Guo, Xuexun; Zuo, Lei

    2016-07-01

    Traditional shock absorbers provide favourable ride comfort and road handling by dissipating the suspension vibration energy into heat waste. In order to harvest this dissipated energy and improve the vehicle fuel efficiency, many energy-harvesting shock absorbers (EHSAs) have been proposed in recent years. Among them, two types of EHSAs have attracted much attention. One is a traditional EHSA which converts the oscillatory vibration into bidirectional rotation using rack-pinion, ball-screw or other mechanisms. The other EHSA is equipped with a mechanical motion rectifier (MMR) that transforms the bidirectional vibration into unidirectional rotation. Hereinafter, they are referred to as NonMMR-EHSA and MMR-EHSA, respectively. This paper compares their performances with the corresponding traditional shock absorber by using closed-form analysis and numerical simulations on various types of vehicles, including passenger cars, buses and trucks. Results suggest that MMR-EHSA provides better ride performances than NonMMR-EHSA, and that MMR-EHSA is able to improve both the ride comfort and road handling simultaneously over the traditional shock absorber when installed on light-damped, heavy-duty vehicles. Additionally, the optimal parameters of MMR-EHSA are obtained for ride comfort. The optimal solutions ('Pareto-optimal solutions') are also obtained by considering the trade-off between ride comfort and road handling.

  19. A Study of the Anechoic Performance of Rice Husk-Based, Geometrically Tapered, Hollow Absorbers

    Directory of Open Access Journals (Sweden)

    Muhammad Nadeem Iqbal

    2014-01-01

    Full Text Available Although solid, geometrically tapered microwave absorbers are preferred due to their better performance, they are bulky and must have a thickness on the order of λ or more. The goal of this study was to design lightweight absorbers that can reduce the electromagnetic reflections to less than −10 dB. We used a very simple approach; two waste materials, that is, rice husks and tire dust in powder form, were used to fabricate two independent samples. We measured and used their dielectric properties to determine and compare the propagation constants and quarter-wave thickness. The quarter-wave thickness for the tire dust was 3 mm less than that of the rice husk material, but we preferred the rice-husk material. This preference was based on the fact that our goal was to achieve minimum backward reflections, and the rice-husk material, with its low dielectric constant, high loss factor, large attenuation per unit length, and ease of fabrication, provided a better opportunity to achieve that goal. The performance of the absorbers was found to be better (lower than −20 dB, and comparison of the results proved that the hollow design with 58% less weight was a good alternative to the use of solid absorbers.

  20. Practical implications of neutron survey instrument performance

    International Nuclear Information System (INIS)

    Tanner, R. J.; Bartlett, D. T.; Hager, I. G.; Jones, I. N.; Molinos, C.; Roberts, N. J.; Taylor, G. C.; Thomas, D. J.

    2004-01-01

    Improvements have been made to the Monte Carlo modelling used to calculate the response of the neutron survey instruments most commonly used in the UK, for neutron energies up to 20 MeV. The improved modelling of the devices includes the electronics and battery pack, allowing better calculations of both the energy and angle dependence of response. These data are used to calculate the response of the instruments in rotationally and fully isotropic, as well as unidirectional fields. Experimental measurements with radionuclide sources and monoenergetic neutron fields have been, and continue to be made, to test the calculated response characteristics. The enhancements to the calculations have involved simulation of the sensitivity of the response to variations in instrument manufacture, and will include the influence of the user and floor during measurements. The practical implications of the energy and angle dependence of response, variations in manufacture, and the influence of the user are assessed by folding the response characteristics with workplace energy and direction distributions. (authors)

  1. Comparison of sound absorbing performances of copper foam and iron foam with the same parameters

    Science.gov (United States)

    Yang, X. C.; Shen, X. M.; Xu, P. J.; Zhang, X. N.; Bai, P. F.; Peng, K.; Yin, Q.; Wang, D.

    2018-01-01

    Sound absorbing performances of the copper foam and the iron foam with the same parameters were investigated by the AWA6128A detector according to standing wave method. Two modes were investigated, which included the pure metal foam mode and the combination mode with the settled thickness of metal foam. In order to legibly compare the sound absorbing coefficients of the two metal foams, the detected sound frequency points were divided into the low frequency range (100 Hz ~ 1000 Hz), the middle frequency range (1000 Hz ~ 3200 Hz), and the high frequency range (3500 Hz ~ 6000 Hz). Sound absorbing performances of the two metal foams in the two modes were discussed within the three frequency ranges in detail. It would be calculated that the average sound absorbing coefficients of copper foam in the pure metal foam mode were 12.6%, 22.7%, 34.6%, 43.6%, 51.1%, and 56.2% when the thickness was 5 mm, 10 mm, 15 mm, 20 mm, 25 mm, and 30 mm. meanwhile, in the combination mode, the average sound absorbing coefficients of copper foam with the thickness of 10 mm were 30.6%, 34.8%, 36.3%, and 35.8% when the cavity was 5 mm, 10 mm, 15 mm, and 20 mm. In addition, those of iron foam in the pure metal foam mode were 13.4%, 20.1%, 34.4%, 43.1%, 49.6%, and 56.1%, and in the combination mode were 25.6%, 30.5%, 34.3%, and 33.4%.

  2. 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.

  3. Thermal Performance Analysis of Staging Effect of Solar Thermal Absorber with Cross Design

    International Nuclear Information System (INIS)

    Amir Abdul Razak; Zafri Azran Abdul Majid; Mohd Hafidz Ruslan; Kamaruzzaman Sopian

    2015-01-01

    The type and shape of solar thermal absorber materials will impact on the operating temperature and thermal energy storage effect of a solar air thermal collector. For a standard flat-plate design, energy gain can be increased by expanding the thermal absorber area along the collector plane, subject to area limitation. This paper focuses on the staging effect of a metal hollow square rod absorber of aluminium, stainless steel, and a combination of the two with a cross design, for the heat gain and temperature characteristics of a solar air collector. Experiments were carried out with three cross design set-ups, with 30 minutes of heating and cooling, phase, respectively, under 485 W/ m 2 solar irradiance value, and at a constant air speed at 0.38 m/ s. One set aluminium set-up delivered the highest output temperature of 41.8 degree Celsius, followed by two-sets aluminium and one aluminium set + one stainless steel set at 39.3 and 38.2 degree Celsius, respectively. The lowest peak temperature is recorded on three sets of the aluminium absorber at 35 degree Celsius. The bi-metallic set-up performed better than the two aluminium set-up where each set-up obtained a temperature drop against heat gain gradient value of -0.4186 degree Celsius/ W and -0.4917 degree Celsius/ W, respectively. Results concluded that by increasing the number of sets, the volume and surface areas of the absorber material are also increased, and lead to a decrease in peak temperature output for each increase of sets. (author)

  4. The Experimental Performance of an Unglazed PVT Collector with Two Different Absorber Types

    Directory of Open Access Journals (Sweden)

    Jin-Hee Kim

    2012-01-01

    Full Text Available Photovoltaic-thermal collectors combine photovoltaic modules and solar thermal collectors, forming a single device that produces electricity and heat simultaneously. There are two types of liquid-type PVT collectors, depending on the existence or absence of a glass cover over the PV module. The glass-covered (glazed PVT collector produces relatively more thermal energy but has a lower electrical yield, whereas the uncovered (unglazed PVT collector has a relatively low thermal energy and somewhat higher electrical performance. The thermal and electrical performance of liquid-type PVT collectors is related not only to the collector design, such as whether a glass cover is used, but also to the absorber design, that is, whether the absorber is for the sheet-and-tube type or the fully wetted type. The design of the absorber, as it comes into contact with the PV modules and the liquid tubes, is regarded as important, as it is related to the heat transfer from the PV modules to the liquid in the tubes. In this paper, the experimental performance of two liquid-type PVT collectors, a sheet-and-tube type and a fully wetted type, was analyzed.

  5. A Field Performance Evaluation Scheme for Microwave-Absorbing Material Coatings

    Directory of Open Access Journals (Sweden)

    Shaopeng Guan

    2017-03-01

    Full Text Available Performance evaluation is an important aspect in the study of microwave-absorbing material coatings. The reflectivity of the incident wave is usually taken as the performance indicator. There have been various methods to directly or indirectly measure the reflectivity, but existing methods are mostly cumbersome and require a strict testing environment. What is more, they cannot be applied to field measurement. In this paper, we propose a scheme to achieve field performance evaluation of microwave-absorbing materials, which adopts a small H-plane sectoral horn antenna as the testing probe and a small microwave reflectometer as the indicator. When the size of the H-plane sectoral horn antenna is specially designed, the field distribution at the antenna aperture can be approximated as a plane wave similar to the far field of the microwave emitted by a radar unit. Therefore, the reflectivity can be obtained by a near-field measurement. We conducted experiments on a kind of ferrite-based microwave-absorbing material at X band (8.2–12.4 GHz to validate the scheme. The experimental results show that the reflectivity is in agreement with the reference data measured by the conventional method as a whole.

  6. How to polarise all neutrons in one beam: a high performance polariser and neutron transport system

    Science.gov (United States)

    Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.

    2016-09-01

    Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.

  7. 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

  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. 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)

  10. 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

  11. High-performance instruments in neutron arena of JHP. Preliminary version

    International Nuclear Information System (INIS)

    Furusaka, M.; Itoh, S.; Otomo, T.; Arai, M.

    1996-05-01

    This report is a preliminary report of high-performance instruments in neutron arena of JHP (Japan Hadron Project). This report consists of as follows; neutron intensity of neutron arena, development of neutron sources in neutron arena, experimental devices and instrumentation. (J.P.N.)

  12. Performance Evaluation of Absorbent Solution for Draw Solute Recovery in Forward Osmosis Desalination Process

    International Nuclear Information System (INIS)

    Kim, Young; Lee, Jong Hoon; Lee, Kong Hoon; Kim, Yu-Chang; Oh, Dong Wook; Lee, Jungho

    2013-01-01

    Although forward osmosis desalination technology has drawn substantial attention as a next-generation desalination method, the energy efficiency of its draw solution treatment process should be improved for its commercialization. When ammonium bicarbonate is used as the draw solute, the system consists of forward-osmosis membrane modules, draw solution separation and recovery processes. Mixed gases of ammonia and carbon dioxide generated during the draws solution separation, need to be recovered to re-concentrate ammonium bicarbonate solution, for continuous operation as well as for the economic feasibility. The diluted ammonium bicarbonate solution has been proposed as the absorbent for the draw solution regeneration. In this study, experiments are conducted to investigate performance and features of the absorption corresponding to absorbent concentration. It is concluded that ammonium bicarbonate solution can be used to recover the generated ammonia and carbon dioxide. The results will be applied to design and operation of pilot-scale forward-osmosis desalination system

  13. Performance Evaluation of Absorbent Solution for Draw Solute Recovery in Forward Osmosis Desalination Process

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young; Lee, Jong Hoon; Lee, Kong Hoon; Kim, Yu-Chang; Oh, Dong Wook; Lee, Jungho [Korea Institute of Machinery Materials, Daejeon (Korea, Republic of)

    2013-04-15

    Although forward osmosis desalination technology has drawn substantial attention as a next-generation desalination method, the energy efficiency of its draw solution treatment process should be improved for its commercialization. When ammonium bicarbonate is used as the draw solute, the system consists of forward-osmosis membrane modules, draw solution separation and recovery processes. Mixed gases of ammonia and carbon dioxide generated during the draws solution separation, need to be recovered to re-concentrate ammonium bicarbonate solution, for continuous operation as well as for the economic feasibility. The diluted ammonium bicarbonate solution has been proposed as the absorbent for the draw solution regeneration. In this study, experiments are conducted to investigate performance and features of the absorption corresponding to absorbent concentration. It is concluded that ammonium bicarbonate solution can be used to recover the generated ammonia and carbon dioxide. The results will be applied to design and operation of pilot-scale forward-osmosis desalination system.

  14. Numerical simulation of the effects of hanging sound absorbers on TABS cooling performance

    DEFF Research Database (Denmark)

    Rage, Nils; Kazanci, Ongun Berk; Olesen, Bjarne W.

    2016-01-01

    simulating a two-person office of 20 m2, with a typical cooling load of 42 W/m2. The results show that covering 60% of the ceiling surface with sound absorbers hanging at 300 mm from the ceiling active deck is expected to reduce the cooling capacity coefficient of TABS by 15.8%. This drops to 25......Recently there has been a considerable increase in the use of Thermally-Active Building Systems (TABS) in Europe as an energy-efficient and economical cooling and heating solution for buildings. However, this widespread solution requires large uncovered hard surfaces indoors, which can lead...... to a degradation of the room acoustic comfort. Therefore, challenges arise when this system has to be combined with acoustic requirements. Soffit-hanging sound absorbers embody a promising solution. This study focuses on quantifying their impact on the cooling performance of TABS, assessed by means of the cooling...

  15. Effect of different absorbing materials on the performance of basin solar still under Libyan climate conditions

    International Nuclear Information System (INIS)

    Shuia, Essaied M.; El-Agouz, Elsayed A.

    2013-01-01

    This experimental study deals with a single-basin solar still using various absorbing materials with and without black painting. Different types of absorbing materials with and without black painting were used to enhance the solar still productivity through improvement in absorptivity. These materials are steel and aluminum with and without black painting and rubber. Two identical solar stills were manufactured using locally available materials. All the results were compared together to reach the best absorbing materials with and without painting that can be used for solar still. it was found that the rubber absorber has the highest water collection during daytime, followed by the black painted steel absorber, then by black painted aluminum absorber and steel without painting absorber. The average enhancement in the daily productivity was about 50% for the rubber absorber compared with the black painted aluminum absorber and about 43% for the rubber absorber compared with the black painted steel absorber.(author)

  16. Parameters Analysis of Hydraulic-Electrical Energy Regenerative Absorber on Suspension Performance

    Directory of Open Access Journals (Sweden)

    Han Zhang

    2014-05-01

    Full Text Available To recycle the vibration energy of vehicles over rough roads, a hydraulic-electricity energy regenerative suspension (HEERS was designed in the present work, and simulations were performed with focus on its performance. On the basis of the system principle, the mathematical model of hydraulic-electrical energy regenerative absorber (HEERA and two degrees of freedom (DOF suspension dynamic model were constructed. Using the model of HEERA, simulations on force-displacement and force-velocity characteristics were performed with a 1.67 Hz frequency and a sinusoidal input adopted. And then in combination with HEERA model and two DOF suspension models, simulations on the performance of HEERS also were carried out. Finally, the influences of charging pressure and volume of the accumulator, hydraulic motor displacement, orifice area of check valve, and inner diameter of hydraulic pipelines on the performance of HEERA and HEERS were investigated in depth. The simulation results indicated that (i the damping characteristic of HEERA was coincident with the damping characteristics of traditional absorber; (ii the most remarkable influencing factor on the performance of HEERS was the hydraulic motor displacement, followed by orifice area of check valve, inner diameter of pipelines, and charging pressure of accumulator, while the effects of charging volume of accumulator were quite limited.

  17. 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)

  18. The development of national quality performance standards for disposable absorbent products for adult incontinence.

    Science.gov (United States)

    Muller, Nancy; McInnis, Elaine

    2013-09-01

    Disposable absorbent products are widely used in inpatient care settings and in the community to manage adult urinary and fecal incontinence, but few product standards exist to help guide their production or optimal use. Increasing costs and reduced revenues have caused a number of states to evaluate absorbent product use among persons who receive care at home with the assistance of the Medicaid Waiver Program, further increasing concerns about the lack of product performance standards. To address these issues, the National Association For Continence (NAFC) formed a council of experts and key stakeholders with the objective of establishing national, independent quality performance standards for disposable absorbent products provided by states to Waiver Program recipients. The Council consisted of representatives from five purposefully selected states, technical directors from six nonwoven product manufacturers, an officer of the nonwoven manufactures trade association, a delegate from an academic nursing program and professional societies, a family caregiver, and a patient representative. Following a consensus method and guidelines for use, nine specific recommendations were developed, posted for public comment, and further refined. Final recommendations for product performance assessment include: rewet rate (a measure of a product's ability to withstand multiple incontinent episodes between changes), rate of acquisition (a measure of the speed at which urine is drawn away from the skin by a product, product retention capacity (a measure of a product's capacity to hold fluid without rewetting the skin), sizing options, absorbency levels, product safety, closure technology, breathable zones (a measure of the air permeability across a textile-like fabric at a controlled differential pressure), and elasticity. The Council also set values for and recommended four quantifiable parameters, and the testing methodology associated with each, to help consumers and states

  19. PERFORMANCE CHARACTERISTICS OF PARABOLIC SOLAR COLLECTOR WATER HEATER SYSTEM FITTED WITH NAIL TWISTED TAPES ABSORBER

    Directory of Open Access Journals (Sweden)

    K. SYED JAFAR

    2017-03-01

    Full Text Available In this paper, the experimental heat transfer, friction loss and thermal performance data for water flowing through the absorber tube fitted with two different twisted tape configurations in parabolic trough collector (PTC are presented. In the present work, a relative experimental study is carried out to investigate the performance of a PTC influenced by heat transfer through fluidabsorber wall mixing mechanism. The major findings of this experiment show that heat transport enhancement in the nail twisted tape collector perform significantly better than plain twisted tapes and also show that the smallest twisted tape ratio enhances the system performance remarkably maximizing the collector efficiency. The results suggest that the twisted tape and nail twisted tape would be a better option for high thermal energy collection in laminar region of the PTC system.

  20. The neutronic design and performance of the Indiana University Cyclotron Facility (IUCF) Low Energy Neutron Source (LENS)

    Science.gov (United States)

    Lavelle, Christopher M.

    Neutron scattering research is performed primarily at large-scale facilities. However, history has shown that smaller scale neutron scattering facilities can play a useful role in education and innovation while performing valuable materials research. This dissertation details the design and experimental validation of the LENS TMR as an example for a small scale accelerator driven neutron source. LENS achieves competitive long wavelength neutron intensities by employing a novel long pulse mode of operation, where the neutron production target is irradiated on a time scale comparable to the emission time of neutrons from the system. Monte Carlo methods have been employed to develop a design for optimal production of long wavelength neutrons from the 9Be(p,n) reaction at proton energies ranging from 7 to 13 MeV proton energy. The neutron spectrum was experimentally measured using time of flight, where it is found that the impact of the long pulse mode on energy resolution can be eliminated at sub-eV neutron energies if the emission time distribution of neutron from the system is known. The emission time distribution from the TMR system is measured using a time focussed crystal analyzer. Emission time of the fundamental cold neutron mode is found to be consistent with Monte Carlo results. The measured thermal neutron spectrum from the water reflector is found to be in agreement with Monte Carlo predictions if the scattering kernels employed are well established. It was found that the scattering kernels currently employed for cryogenic methane are inadequate for accurate prediction of the cold neutron intensity from the system. The TMR and neutronic modeling have been well characterized and the source design is flexible, such that it is possible for LENS to serve as an effective test bed for future work in neutronic development. Suggestions for improvements to the design that would allow increased neutron flux into the instruments are provided.

  1. Mass transfer performance of blended alkanolamines for CO{sub 2} capture in packed absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Setameteekul, A.; Veawab, A.; Aroonwilas, A.; Tontiwachwuthikul, P. [Regina Univ., SK (Canada)

    2003-07-01

    Acid gases are removed from industrial gas streams using the alkanolamine absorption process. There has been recent interest in extending the process to remove carbon dioxide from industrial waste gases. The process based on conventional alkanolamines is not economically viable because of the associated high energy costs. It was suggested that blended alkanolamines would significantly reduce energy consumption, thereby resulting in a reduction in process costs. The main disadvantage of using blended alkanolamines is a decrease in absorption performance. This study examines the mass transfer behaviour of carbon dioxide into blended alkanolamine solutions. It also compares their performance with the baseline performance of monoethanolamine (MEA). A series of absorption experiments were conducted in a bench-scale packed absorber. The blended alkanolamines included mixtures of MEA and methyldiethanolamine, as well as mixtures of diethanolamine and methyldiethanolamine. The results indicated the general mass transfer coefficient as a function of operating conditions.

  2. An easier method for performing a pancreaticojejunostomy for the soft pancreas using a fast-absorbable suture

    OpenAIRE

    Hakamada, Kenichi; Narumi, Shunji; Toyoki, Yoshikazu; Nara, Masaki; Ishido, Kenosuke; Miura, Takuya; Kubo, Norihito; Sasaki, Mutsuo

    2008-01-01

    AIM: To clarify the usefulness of a new method for performing a pancreaticojejunostomy by using a fast-absorbable suture material irradiated polyglactin 910, and a temporary stent tube for a narrow pancreatic duct with a soft pancreatic texture.

  3. Experimental and Numerical Study of the Effects of Acoustic Sound Absorbers on the Cooling Performance of Thermally Active Building Systems

    DEFF Research Database (Denmark)

    Domínguez, L. Marcos; Kazanci, Ongun Berk; Rage, Nils

    2017-01-01

    Free-hanging horizontal and vertical sound absorbers are commonly used in buildings for room acoustic control; however, when these sound absorbers are used in combination with Thermally Active Building Systems, they will decrease the cooling performance of Thermally Active Building Systems...... and this will affect the thermal indoor environment in that space. Therefore, it is crucial to be able to quantify and model these effects in the design phase. This study quantifies experimentally the effects of horizontal and vertical free-hanging sound absorbers on the cooling performance of Thermally Active......%, respectively. With vertical sound absorbers, the decrease in cooling performance was 8%, 12%, and 14% for the corresponding cases, respectively. The numerical model predicted closely the cooling performance reduction, air temperatures and ceiling surface temperatures in most cases, while there were differences...

  4. Enhancing the Performance of the Microwave Absorbing Materials by Using Dielectric Resonator Arrays

    Directory of Open Access Journals (Sweden)

    Omar H. Al-Zoubi

    2017-01-01

    Full Text Available We present a technique for enhancing the performance of microwave absorbing materials in terms of weight, thickness, and bandwidth. The introduced technique is based on fabricating the microwave absorbing (MA material in a structure comprised of an array of circular cylinder dielectric resonators (CDR backed by a perfect electric conductor (PEC ground plane. Numerical electromagnetic methods are employed to study the properties of the proposed MA array structures, where 3D full wave simulation using finite-element method is implemented. The obtained results show that the performance of the MA-CDR arrays significantly outperforms that of a flat layer composed of the same material and having equivalent thickness. A flat layer of MA material with thickness of 5 mm backed by perfect electric conductor (PEC shows as low as -50 dB reflection loss (RL peak and ~3 GHz 10-dB bandwidth, whereas an MA-CDR array, composed of the same MA material, of height of 4 mm can achieve as low as ~−50 dB RL peak and ~12 GHz 10-dB RL bandwidth.

  5. The EUROBALL neutron wall - design and performance tests of neutron detectors

    CERN Document Server

    Skeppstedt, Ö; Lindström, L; Wadsworth, R; Hibbert, I; Kelsall, N; Jenkins, D; Grawe, H; aGórska, M; Moszynski, M; Sujkowski, Z; Wolski, D; Kapusta, M; Hellström, M; Kalogeropoulos, S; Oner, D; Johnson, A; Cederkäll, J; Klamra, W; Nyberg, J; Weiszflog, M; Kay, J; Griffiths, R; Garces-Narro, J; Pearson, C; Eberth, J

    1999-01-01

    The mechanical design of the EUROBALL neutron wall and neutron detectors, and their performance measured with a sup 2 sup 4 sup 6 sup , sup 2 sup 4 sup 8 Cm fission source are described. The array consists of 15 pseudohexaconical detector units subdivided into three, 149 mm high, hermetically separated segments and a smaller central pentagonal unit subdivided into five segments. The detectors are filled with Bicron BC501A liquid scintillator. Each section of the hexaconical detectors is viewed by a 130 mm diameter Philips XP4512PA photomultiplier while the sections of pentagonal detectors are viewed by Philips XP4312B PMTs. The tests of n-gamma discrimination performed by zero-crossing and time-of-flight methods show a full separation of gamma- and neutron events down to 50 keV recoil electron energy. These tests demonstrate the excellent timing properties of the detectors and an average time resolution of 1.56 ns. The factors determining the efficiency of neutron detectors are discussed. The total efficiency...

  6. Postirradiation examination of JOYO MK-II control rod (CRM601). Irradiation performance of shroud type absorber pin

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Kikuchi, Shin; Katsuyama, Kozo; Nagamine, Tsuyoshi; Mitsugi, Takeshi; Uto, Manabu; Tatebe, Kazuaki; Onose, Shoji; Maruyama, Tadashi

    1998-10-01

    This paper describes the results of postirradiation examination and analysis by CORAL code for irradiation performance of CRM601 control rod, which was the 6th reloaded control rod with shroud type absorber pins for use in JOYO MK-II core. The detailed visual examination indicated that there was no cladding breach in absorber pins. However, sodium ingress from the vent tube was observed in four absorber pins among seven pins. While a remarkable oval deformation occurred in cladding tube of helium bonded absorber pins, a little or no diametral change was observed in the absorber pins in which sodium ingress took place. From metallurgical observations and the analysis by CORAL code, it was estimated that the shroud tube installed in helium bonded absorber pins were irradiated at 720degC, and those in sodium bonded absorber pins were irradiated at 420degC. It was confirmed that diametral change of cladding depended on the initial gap between shroud and cladding tube. The results of present investigation indicate that it is desirable to use the materials with low thermal expansion coefficient for shroud tubes, and that sodium bonded absorber pins were advantageous for obtaining long life control rods. (author)

  7. Performance advantages of CPML over UPML absorbing boundary conditions in FDTD algorithm

    Science.gov (United States)

    Gvozdic, Branko D.; Djurdjevic, Dusan Z.

    2017-01-01

    Implementation of absorbing boundary condition (ABC) has a very important role in simulation performance and accuracy in finite difference time domain (FDTD) method. The perfectly matched layer (PML) is the most efficient type of ABC. The aim of this paper is to give detailed insight in and discussion of boundary conditions and hence to simplify the choice of PML used for termination of computational domain in FDTD method. In particular, we demonstrate that using the convolutional PML (CPML) has significant advantages in terms of implementation in FDTD method and reducing computer resources than using uniaxial PML (UPML). An extensive number of numerical experiments has been performed and results have shown that CPML is more efficient in electromagnetic waves absorption. Numerical code is prepared, several problems are analyzed and relative error is calculated and presented.

  8. Neutron activation studies on JET

    International Nuclear Information System (INIS)

    Loughlin, M.J.; Forrest, R.A.; Edwards, J.E.G.

    2001-01-01

    Extensive neutron transport calculations have been performed to determine the neutron spectrum at a number of points throughout the JET torus hall. The model has been bench-marked against a set of foil activation measurements which were activated during an experimental campaign with deuterium/tritium plasmas. The model can predict the neutron activation of the foils on the torus hall walls to within a factor of three for reactions with little sensitivity to thermal neutrons. The use of scandium foils with and without a cadmium thermal neutron absorber was a useful monitor of the thermal neutron flux. Conclusions regarding the usefulness of other foils for benchmarking the calculations are also given

  9. Nano-hydroxyapatite as an Efficient Polysulfide Absorbent for High-performance Li-S Batteries

    International Nuclear Information System (INIS)

    Liu, Naiqiang; Ai, Fei; Wang, Weikun; Shao, Hongyuan; Zhang, Hao; Wang, Anbang; Xu, Zhichuan J.; Huang, Yaqin

    2016-01-01

    Highlights: • Nano-HA has been demonstrated as an efficient polysulfide absorbent. • The shuttle effect of polysulfide in Li-S battery has been confined by the nano-HA. • Nano-HA used as additive improved electrochemical performance of Li-S battery. - Abstract: Lithium-sulfur (Li-S) battery is regarded as one of the most promising candidates for developing advanced energy storage system, but the polysulfide shuttle effect remains the biggest obstacle for its practical application. In this work, nano-hydroxyapatite (Ca 5 (PO 4 ) 3 (OH)) was used as an additive in the sulfur cathode and carbon-coated separator to prevent the polysulfide shuttle effect and thus to achieve the high performance. The sulfur cathode with nano-hydroxyapatite exhibited a higher reversible capacity and a more stable cycling performance than that of the pristine sulfur cathode. The improved capacity retention from 58% (100th) to 73% (200th) after introducing nano-hydroxyapatite into the sulfur cathode confirmed its strong polysulfide absorption ability. Furthermore, a nano-hydroxyapatite modified separator was developed to suppress the polysulfide shuttle effect and to facilitate the reutilization of sulfur species. The nano-hydroxyapatite particles served as polysulfide absorbents to bind polysulfides and suppress their diffusion to the anode. The batteries assembled with this separator exhibited a high reversible capacity of 886 mAhg −1 at 0.1C and 718 mAh g −1 at 0.5C after 200 cycles, with a low capacity fading of ∼0.10-0.11% per-cycle. At the highest sulfur loading of 4.5 mg cm −2 used for practical applications, the reversible areal capacity was much higher than the areal capacity (4 mAh cm −2 ) of commercial lithium-ion batteries. Therefore, the strategy using nano-hydroxyapatite as polysulfide absorbent shows great potential for solving the polysulfide shuttle problem and developing high performance Li-S batteries.

  10. A Solar Volumetric Receiver: Influence of Absorbing Cells Configuration on Device Thermal Performance

    Science.gov (United States)

    Yilbas, B. S.; Shuja, S. Z.

    2017-01-01

    Thermal performance of a solar volumetric receiver incorporating the different cell geometric configurations is investigated. Triangular, hexagonal, and rectangular absorbing cells are incorporated in the analysis. The fluid volume fraction, which is the ratio of the volume of the working fluid over the total volume of solar volumetric receiver, is introduced to assess the effect of cell size on the heat transfer rates in the receiver. In this case, reducing the fluid volume fraction corresponds to increasing cell size in the receiver. SiC is considered as the cell material, and air is used as the working fluid in the receiver. The Lambert's Beer law is incorporated to account for the solar absorption in the receiver. A finite element method is used to solve the governing equation of flow and heat transfer. It is found that the fluid volume fraction has significant effect on the flow field in the solar volumetric receiver, which also modifies thermal field in the working fluid. The triangular absorbing cell gives rise to improved effectiveness of the receiver and then follows the hexagonal and rectangular cells. The second law efficiency of the receiver remains high when hexagonal cells are used. This occurs for the fluid volume fraction ratio of 0.5.

  11. LANSCE (Los Alamos Neutron Scattering Center) target system performance

    International Nuclear Information System (INIS)

    Russell, G.J.; Gilmore, J.S.; Robinson, H.; Legate, G.L.; Bridge, A.; Sanchez, R.J.; Brewton, R.J.; Woods, R.; Hughes, H.G. III

    1989-01-01

    The authors measured neutron beam fluxes at LANSCE using gold foil activation techniques. They did an extensive computer simulation of the as-built LANSCE Target/Moderator/Reflector/Shield geometry. They used this mockup in a Monte Carlo calculation to predict LANSCE neutronic performance for comparison with measured results. For neutron beam fluxes at 1 eV, the ratio of measured data to calculated varies from ∼0.6-0.9. The computed 1 eV neutron leakage at the moderator surface is 3.9 x 10 10 n/eV-sr-s-μA for LANSCE high-intensity water moderators. The corresponding values for the LANSCE high-resolution water moderator and the liquid hydrogen moderator are 3.3 and 2.9 x 10 10 , respectively. LANSCE predicted moderator intensities (per proton) for a tungsten target are essentially the same as ISIS predicted moderator intensities for a depleted uranium target. The calculated LANSCE steady state unperturbed thermal (E 13 n/cm 2 -s. The unique LANSCE split-target/flux-trap-moderator system is performing exceedingly well. The system has operated without a target or moderator change for over three years at nominal proton currents of 25 μA of 800-MeV protons. 17 refs., 8 figs., 3 tabs

  12. Neutronic performance of a benchmark 1-MW LPSS

    International Nuclear Information System (INIS)

    Russell, G.J.; Pitcher, E.J.; Ferguson, P.D.

    1995-01-01

    We used split-target/flux-trap-moderator geometry in our 1-MW LPSS computational benchmark performance calculations because the simulation models were readily available. Also, this target/moderator arrangement is a proven LANSCE design and a good neutronic performer. The model has four moderator viewed surfaces, each with a 13x13 cm field-of-view. For our scoping neutronic-performance calculations, we attempted to get as much engineering realism into the target-system mockup as possible. In our present model, we account for target/reflector dilution by cooling; the D 2 O coolant fractions are adequate for 1 MW of 800-MeV protons (1.25 mA). We have incorporated a proton beam entry window and target canisters into the model, as well as (partial) moderator and vacuum canisters. The model does not account for target and moderator cooling lines and baffles, entire moderator canisters, and structural material in the reflector

  13. Performance following a 500-675 rad neutron pulse

    International Nuclear Information System (INIS)

    Yochmowitz, M.G.; Brown, G.C.; Hardy, K.A.

    1985-01-01

    A three-light, three-lever discrete avoidance behavioral task was initiated to study the effects of a 500-675 rad neutron pulse upon performance. Eight primates performed the task for 4 h (3.5 h postexposure) on exposure day and for 4 h on each of 3 d postexposure. For the exposure day, five subjects had a decrease in correct responses, seven had increased reaction times, and six experienced productive emesis within 3.5 hours postexposure. Although the performance degradations were not severe, these data suggest that the performance of time critical tasks could be significantly impaired. 10 references

  14. Study on Ballistic Absorbing Energy Character of High Performance Polyethylene Needle Felt

    Science.gov (United States)

    Kailiang, Zhu; Jianqiao, Fu

    2017-11-01

    The ballistic performance of polyethylene needle felt is tested and the failure morphology after test is also observed. The results showed that when the non-dimensionally non-stressed fibers in polyethylene needles are subjected to high-speed projectile, secondary movement such as stretching and twisting occurs first. This secondary movement is very full, it is the main way of ballistic absorbing energy of the polyethylene needle felt which can avoid the polyethylene fiber short-term rapid heating-up and destroyed. Analysis results show that under normal temperature and humidity conditions, the V50 of 6-layer forded polyethylene needle felt sample is 250m/s. At (450 ± 50) m/s speed range of the target missile, the mean value of the penetrative specific energy absorption for 3-layer forded polyethylene needle felt anti-1.1g simulated projectiles (tapered column) reaches 24.1J·m2/kg.

  15. Performance study of solar air heater duct having absorber plate with V down perforated baffles

    Directory of Open Access Journals (Sweden)

    Sunil Chamoli

    2014-04-01

    Full Text Available This paper presents results of a study of the performance of solar air heaters with V down perforated baffles as roughness on the air flow side of the absorber plate. Investigations have been carried out using a mathematical model to study the effects of ambient conditions, operating and design parameters on effective efficiency of such air heaters. The thermal and effective efficiencies differ only marginally at lower flow rates. With an increase in the flow rate, the difference between the thermal and effective efficiencies increases because of the increase in the pumping power. The results of the study are presented in the form of plots to show the effect of ambient, design and operating conditions on thermal and effective efficiency.

  16. 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)

  17. Regeneration performance of CO2-rich solvents by using membrane vacuum regeneration technology: Relationships between absorbent structure and regeneration efficiency

    International Nuclear Information System (INIS)

    Yan, Shuiping; Fang, Mengxiang; Wang, Zhen; Luo, Zhongyang

    2012-01-01

    Highlights: ► MVR may be viable to successfully use less valuable heat to replace high grade steam. ► Increasing OH and amine groups will increase the regeneration efficiency. ► Absorbents with a four carbon chain length will be more attractive to MVR. ► Amino acid salts will be more appropriate for MVR. ► HRM conducted at ambient pressure and low temperature is inferior to MVR. -- Abstract: In order to give a better understanding for the selection of suitable absorbents for the novel membrane vacuum regeneration technology (MVR) which has the potential to reduce CO 2 energy requirement by utilizing the waste heat or low-grade energy, an experimental study to determine the relationships between chemical structure and vacuum regeneration behavior of CO 2 absorbents at 70 °C and 10 kPa was performed. Eleven typical absorbents with different functional groups in their chemical structures were investigated in terms of vacuum regeneration efficiencies. Results showed that the regeneration efficiency decreased with an increase of number of activated hydrogen atom in amine group and decreased with the number of hydroxyl group. Especially, more attention should be paid to these alkanolamines with one hydrogen atom in amine group and two or more hydroxyl groups in the structures due to their better comprehensive performance in regeneration, absorbent loss and CO 2 absorption aspects. Increasing the carbon chain length and amine groups in the absorbent structure contributed to the improvement of regeneration performance and reduction of absorbent volatile loss. These absorbents with a four carbon chain length bonded at amine group might be more attractive to MVR. Furthermore, polyamines were superior to monoamines in terms of higher regeneration efficiencies and lower absorbent losses. Additionally, the individual effects of the potassium carboxylate group and hydroxymethylene group were also compared in this study. Results showed that amino acid salts were more

  18. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  19. Systematic study on the performance of elliptic focusing neutron guides

    International Nuclear Information System (INIS)

    Martin Rodriguez, D.; DiJulio, D.D.; Bentley, P.M.

    2016-01-01

    In neutron scattering experiments there is an increasing trend towards the study of smaller volume samples, which make the use of focusing optics more important. Focusing guide geometries based on conic-sections, such as those with parabolic and elliptic shapes, have been extensively used in both recently built neutron instruments and upgrades of existing hardware. A large fraction of proposed instruments at the European Spallation Source feature the requirement of good performance when measuring on small samples. The optimised design of a focusing system comes after time consuming Monte-Carlo (MC) simulations. Therefore, in order to help reduce the time needed to design such focusing systems, it is necessary to study systematically the performance of focusing guides. In the present work, we perform a theoretical analysis of the focusing properties of neutron beams, and validate them using a combination of Monte-Carlo simulations and Particle Swarm Optimisations (PSOs), where there is a close correspondence between the maximum divergence of the beam and the shape of the guide. The analytical results show that two limits can be considered, which bound a range of conic section shapes that provide optimum performance. Finally, we analyse a more realistic guide example and we give an assessment of the importance of the contribution from multiple reflections in different systems.

  20. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

  1. 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

  2. Performance of fast-absorbable suture and histo-glue in closing incisions in Brown trout

    DEFF Research Database (Denmark)

    Jepsen, Niels; Larsen, Martin Hage; Aarestrup, Kim

    2017-01-01

    , growth, tag expulsion rate and incision healing was compared among three groups of dummy transmitter-tagged wild brown trout Salmo trutta where incisions were closed with two types of suture material (absorbable vs. fast absorbable) and Histo-glue. The tagged fish were kept in semi-natural ponds for 20...

  3. Performance of non-conventional factorization approaches for neutron kinetics

    International Nuclear Information System (INIS)

    Bulla, S.; Nervo, M.

    2013-01-01

    The use of factorization techniques provides a interesting option for the simulation of the time-dependent behavior of nuclear systems with a reduced computational effort. While point kinetics neglects all spatial and spectral effects, quasi-statics and multipoint kinetics allow to produce results with a higher accuracy for transients involving relevant modifications of the neutron distribution. However, in some conditions these methods can not work efficiently. In this paper, we discuss some possible alternative formulations for the factorization process for neutron kinetics, leading to mathematical models of reduced complications that can allow an accurate simulation of transients involving spatial and spectral effects. The performance of these innovative approaches are compared to standard techniques for some test cases, showing the benefits and shortcomings of the method proposed. (authors)

  4. Effect of blanket assembly shuffling on LMR neutronic performance

    International Nuclear Information System (INIS)

    Khalil, H.; Fujita, E.K.

    1987-01-01

    Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

  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. A Simulation Study on the Performance of Radiant Ceilings Combined with Free-Hanging Horizontal Sound Absorbers

    DEFF Research Database (Denmark)

    Kazanci, Ongun Berk; Domínguez, L. Marcos; Rage, Niels

    2018-01-01

    using TABS, most building simulation models assume an uncovered ceiling; however, this might not be the case in practice, due to the use of free-hanging horizontal (or vertical) sound absorbers for the control of room acoustic conditions. The use of sound absorbers will decrease the performance...... of radiant ceiling cooling systems. Therefore, the quantification of the effects during the design phase is important for predicting the resulting thermal indoor environment and for system dimensioning. In this study, a two-person office room equipped with TABS was simulated using a commercially available...... simulation software with a recently developed plug-in that allows simulating the effects of horizontal sound absorbers on the performance of TABS and on the thermal indoor environment. The change in thermal indoor environment and in performance of TABS were quantified, and the simulation results were...

  7. 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.

  8. A Long-Pulse Spallation Source at Los Alamos: Facility description and preliminary neutronic performance for cold neutrons

    International Nuclear Information System (INIS)

    Russell, G.J.; Weinacht, D.J.; Pitcher, E.J.; Ferguson, P.D.

    1998-03-01

    The Los Alamos National Laboratory has discussed installing a new 1-MW spallation neutron target station in an existing building at the end of its 800-MeV proton linear accelerator. Because the accelerator provides pulses of protons each about 1 msec in duration, the new source would be a Long Pulse Spallation Source (LPSS). The facility would employ vertical extraction of moderators and reflectors, and horizontal extraction of the spallation target. An LPSS uses coupled moderators rather than decoupled ones. There are potential gains of about a factor of 6 to 7 in the time-averaged neutron brightness for cold-neutron production from a coupled liquid H 2 moderator compared to a decoupled one. However, these gains come at the expense of putting ''tails'' on the neutron pulses. The particulars of the neutron pulses from a moderator (e.g., energy-dependent rise times, peak intensities, pulse widths, and decay constant(s) of the tails) are crucial parameters for designing instruments and estimating their performance at an LPSS. Tungsten is the reference target material. Inconel 718 is the reference target canister and proton beam window material, with Al-6061 being the choice for the liquid H 2 moderator canister and vacuum container. A 1-MW LPSS would have world-class neutronic performance. The authors describe the proposed Los Alamos LPSS facility, and show that, for cold neutrons, the calculated time-averaged neutronic performance of a liquid H 2 moderator at the 1-MW LPSS is equivalent to about 1/4th the calculated neutronic performance of the best liquid D 2 moderator at the Institute Laue-Langevin reactor. They show that the time-averaged moderator neutronic brightness increases as the size of the moderator gets smaller

  9. Performance modeling of parallel algorithms for solving neutron diffusion problems

    International Nuclear Information System (INIS)

    Azmy, Y.Y.; Kirk, B.L.

    1995-01-01

    Neutron diffusion calculations are the most common computational methods used in the design, analysis, and operation of nuclear reactors and related activities. Here, mathematical performance models are developed for the parallel algorithm used to solve the neutron diffusion equation on message passing and shared memory multiprocessors represented by the Intel iPSC/860 and the Sequent Balance 8000, respectively. The performance models are validated through several test problems, and these models are used to estimate the performance of each of the two considered architectures in situations typical of practical applications, such as fine meshes and a large number of participating processors. While message passing computers are capable of producing speedup, the parallel efficiency deteriorates rapidly as the number of processors increases. Furthermore, the speedup fails to improve appreciably for massively parallel computers so that only small- to medium-sized message passing multiprocessors offer a reasonable platform for this algorithm. In contrast, the performance model for the shared memory architecture predicts very high efficiency over a wide range of number of processors reasonable for this architecture. Furthermore, the model efficiency of the Sequent remains superior to that of the hypercube if its model parameters are adjusted to make its processors as fast as those of the iPSC/860. It is concluded that shared memory computers are better suited for this parallel algorithm than message passing computers

  10. Performing Neutron Cross-Section Measurements at RIA

    International Nuclear Information System (INIS)

    Ahle, L.E.

    2003-01-01

    The Rare Isotope Accelerator (RIA) is a proposed accelerator for the low energy nuclear physics community. Its goal is to understand the natural abundances of the elements heavier than iron, explore the nuclear force in systems far from stability, and study symmetry violation and fundamental physics in nuclei. To achieve these scientific goals, RIA promises to produce isotopes far from stability in sufficient quantities to allow experiments. It would also produce near stability isotopes at never before seen production rates, as much as 10 12 pps. Included in these isotopes are many that are important to stockpile stewardship, such as 87 Y, 146-50 Eu, and 231 Th. Given the expected production rates at RIA and a reasonably intense neutron source, one can expect to make ∼10 μg targets of nuclei with a half-life of ∼1 day. Thus, it will be possible at RIA to obtain experimental information on the neutron cross section for isotopes that have to date only been determined by theory. There are two methods to perform neutron cross-section measurements, prompt and delayed. The prompt method tries to measure each reaction as it happens. The exact technique employed will depend on the reaction of interest, (n,2n), (n,γ), (n,p), etc. The biggest challenge with this method is designing a detector system that can handle the gamma ray background from the target. The delayed method, which is the traditional radiochemistry method for determining the cross-section, irradiates the targets and then counts the reaction products after the fact. While this allows one to avoid the target background, the allowed fraction of target impurities is extremely low. This is especially true for the desired reaction product with the required impurity fraction on the order of 10 -9 . This is particularly problematic for (n,2n) and (n,γ) reactions, whose reaction production cannot be chemically separated from the target. In either case, the first step at RIA to doing these measurements is

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. Analysis of sound absorption performance of an electroacoustic absorber using a vented enclosure

    Science.gov (United States)

    Cho, Youngeun; Wang, Semyung; Hyun, Jaeyub; Oh, Seungjae; Goo, Seongyeol

    2018-03-01

    The sound absorption performance of an electroacoustic absorber (EA) is primarily influenced by the dynamic characteristics of the loudspeaker that acts as the actuator of the EA system. Therefore, the sound absorption performance of the EA is maximum at the resonance frequency of the loudspeaker and tends to degrade in the low-frequency and high-frequency bands based on this resonance frequency. In this study, to adjust the sound absorption performance of the EA system in the low-frequency band of approximately 20-80 Hz, an EA system using a vented enclosure that has previously been used to enhance the radiating sound pressure of a loudspeaker in the low-frequency band, is proposed. To verify the usefulness of the proposed system, two acoustic environments are considered. In the first acoustic environment, the vent of the vented enclosure is connected to an external sound field that is distinct from the sound field coupled to the EA. In this case, the acoustic effect of the vented enclosure on the performance of the EA is analyzed through an analytical approach using dynamic equations and an impedance-based equivalent circuit. Then, it is verified through numerical and experimental approaches. Next, in the second acoustic environment, the vent is connected to the same external sound field as the EA. In this case, the effect of the vented enclosure on the EA is investigated through an analytical approach and finally verified through a numerical approach. As a result, it is confirmed that the characteristics of the sound absorption performances of the proposed EA system using the vented enclosure in the two acoustic environments considered in this study are different from each other in the low-frequency band of approximately 20-80 Hz. Furthermore, several case studies on the change tendency of the performance of the EA using the vented enclosure according to the critical design factors or vent number for the vented enclosure are also investigated. In the future

  13. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    International Nuclear Information System (INIS)

    Papukchiev, Angel; Schaefer, Anselm

    2008-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  14. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  15. Design and performance simulation of a segmented-absorber based muon detection system for high energy heavy ion collision experiments

    International Nuclear Information System (INIS)

    Ahmad, S.; Bhaduri, P.P.; Jahan, H.; Senger, A.; Adak, R.; Samanta, S.; Prakash, A.; Dey, K.; Lebedev, A.; Kryshen, E.; Chattopadhyay, S.; Senger, P.; Bhattacharjee, B.; Ghosh, S.K.; Raha, S.; Irfan, M.; Ahmad, N.; Farooq, M.; Singh, B.

    2015-01-01

    A muon detection system (MUCH) based on a novel concept using a segmented and instrumented absorber has been designed for high-energy heavy-ion collision experiments. The system consists of 6 hadron absorber blocks and 6 tracking detector triplets. Behind each absorber block a detector triplet is located which measures the tracks of charged particles traversing the absorber. The performance of such a system has been simulated for the CBM experiment at FAIR (Germany) that is scheduled to start taking data in heavy ion collisions in the beam energy range of 6–45 A GeV from 2019. The muon detection system is mounted downstream to a Silicon Tracking System (STS) that is located in a large aperture dipole magnet which provides momentum information of the charged particle tracks. The reconstructed tracks from the STS are to be matched to the hits measured by the muon detector triplets behind the absorber segments. This method allows the identification of muon tracks over a broad range of momenta including tracks of soft muons which do not pass through all the absorber layers. Pairs of oppositely charged muons identified by MUCH could therefore be combined to measure the invariant masses in a wide range starting from low mass vector mesons (LMVM) up to charmonia. The properties of the absorber (material, thickness, position) and of the tracking chambers (granularity, geometry) have been varied in simulations of heavy-ion collision events generated with the UrQMD generator and propagated through the setup using the GEANT3, the particle transport code. The tracks are reconstructed by a Cellular Automaton algorithm followed by a Kalman Filter. The simulations demonstrate that low mass vector mesons and charmonia can be clearly identified in central Au+Au collisions at beam energies provided by the international Facility for Antiproton and Ion Research (FAIR)

  16. Performance of neutron kinetics models for ADS transient analyses

    International Nuclear Information System (INIS)

    Rineiski, A.; Maschek, W.; Rimpault, G.

    2002-01-01

    Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One

  17. Massively parallel performance of neutron transport response matrix algorithms

    International Nuclear Information System (INIS)

    Hanebutte, U.R.; Lewis, E.E.

    1993-01-01

    Massively parallel red/black response matrix algorithms for the solution of within-group neutron transport problems are implemented on the Connection Machines-2, 200 and 5. The response matrices are dericed from the diamond-differences and linear-linear nodal discrete ordinate and variational nodal P 3 approximations. The unaccelerated performance of the iterative procedure is examined relative to the maximum rated performances of the machines. The effects of processor partitions size, of virtual processor ratio and of problems size are examined in detail. For the red/black algorithm, the ratio of inter-node communication to computing times is found to be quite small, normally of the order of ten percent or less. Performance increases with problems size and with virtual processor ratio, within the memeory per physical processor limitation. Algorithm adaptation to courser grain machines is straight-forward, with total computing time being virtually inversely proportional to the number of physical processors. (orig.)

  18. Simultaneous determination of some ultraviolet-absorbing chemicals in sunscreen cosmetics using a high-performance liquid chromatography method.

    Science.gov (United States)

    Liu, T; Wu, D

    2011-10-01

    A method of gradient elution high-performance liquid chromatography (HPLC) for simultaneous determination of 11 different ultraviolet-absorbing chemicals of phenylbenzlmldazole sulphonic acid, 4-aminobenzoic acid, benzophenone-4, benzophenone-3, isoamyl p-methoxycinnamate, 4-methylbenzylidene camphor, octocrylene, ethylhexyl methoxycinnamate, homosalate, ethylhexyl salicylate, methylene bis-benzotriazolyl tetramethylbutyl phenol was developed for the application to sunscreen cosmetic products. In this study, an Agilent SB-C18 analytical column (250 × 4.6 mm, 5 μm) was utilized and methanol, tetrahydrofuran and perchloric acid aqueous solution (0.2 mL HClO(4) + 300 mL H(2)O) were used for gradient elution at a total flow rate of 1.0 mL min(-1). The optimum conditions for 11 different ultraviolet-absorbing chemicals analyses were investigated. All calibration curves showed good linear regression with UV detection (311 nm) within test ranges. The correlation coefficients were better than 0.999 in all cases. The assay was simple, selective, convenient and reproducible and is suitable for the determination of ultraviolet-absorbing chemicals in commercial sunscreen cosmetic products. The use frequency of 11 different ultraviolet absorbents in 100 sunscreen cosmetics was investigated and statistically analysed. The ultraviolet absorbent of maximum use frequency was ethylhexyl methoxycinnamate. © 2011 The Authors. ICS © 2011 Society of Cosmetic Scientists and the Société Française de Cosmétologie.

  19. Influence of the local absorber layer thickness on the performance of ZnO nanorod solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Belaidi, Abdelhak; Dittrich, Thomas; Kieven, David; Tornow, Julian; Schwarzburg, Klaus; Lux-Steiner, Martha [Helmholtz-Zentrum Berlin fuer Materialien und Energie, Berlin (Germany)

    2008-08-15

    The local absorber layer thickness (d{sub local}) of solar cells with extremely thin absorber was changed between 10 nm and 70 nm. As a model system, ZnO nanorod arrays (electron conductor) with fixed internal surface area coated with In{sub 2}S{sub 3} (absorber) and impregnated with CuSCN (transparent hole conductor) were applied. The performance of the small area solar cells depended critically on d{sub local}. The highest short circuit current density was reached for the lowest d{sub local}. In contrast, the highest open circuit voltage was obtained for the highest d{sub local}. A maximum energy conversion efficiency of 3.4% at AM1.5 was achieved. Limiting factors are discussed.(copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  20. Neutronic calculations for JET. Performed with the FURNACE2 program. (Final report JET contract JEO/9004)

    International Nuclear Information System (INIS)

    Verschuur, K.A.

    1996-10-01

    Neutron-transport calculations with the FURNACE(2) program system, in support of the Neutron Diagnostic Group at JET, have been performed since 1980, i.e. since the construction phase of JET. FURNACE(2) is a ray-tracing/multiple-reflection transport program system for toroidal geometries, that orginally was developed for blanket neutronics studies and which then was improved and extended for application to the neutron-diagnostics at JET. (orig./WL)

  1. Performance of a new carbon dioxide absorbent, Yabashi lime® as compared to conventional carbon dioxide absorbent during sevoflurane anesthesia in dogs.

    Science.gov (United States)

    Kondoh, Kei; Atiba, Ayman; Nagase, Kiyoshi; Ogawa, Shizuko; Miwa, Takashi; Katsumata, Teruya; Ueno, Hiroshi; Uzuka, Yuji

    2015-08-01

    In the present study, we compare a new carbon dioxide (CO2) absorbent, Yabashi lime(®) with a conventional CO2 absorbent, Sodasorb(®) as a control CO2 absorbent for Compound A (CA) and Carbon monoxide (CO) productions. Four dogs were anesthetized with sevoflurane. Each dog was anesthetized with four preparations, Yabashi lime(®) with high or low-flow rate of oxygen and control CO2 absorbent with high or low-flow rate. CA and CO concentrations in the anesthetic circuit, canister temperature and carbooxyhemoglobin (COHb) concentration in the blood were measured. Yabashi lime(®) did not produce CA. Control CO2 absorbent generated CA, and its concentration was significantly higher in low-flow rate than a high-flow rate. CO was generated only in low-flow rate groups, but there was no significance between Yabashi lime(®) groups and control CO2 absorbent groups. However, the CO concentration in the circuit could not be detected (≤5ppm), and no change was found in COHb level. Canister temperature was significantly higher in low-flow rate groups than high-flow rate groups. Furthermore, in low-flow rate groups, the lower layer of canister temperature in control CO2 absorbent group was significantly higher than Yabashi lime(®) group. CA and CO productions are thought to be related to the composition of CO2 absorbent, flow rate and canister temperature. Though CO concentration is equal, it might be safer to use Yabashi lime(®) with sevoflurane anesthesia in dogs than conventional CO2 absorbent at the point of CA production.

  2. Evaluation of Solar Air Heater Performance with Artificial Rib Roughness over the Absorber Plate using Finite Element Modelling Analysis

    Science.gov (United States)

    Kumar, K. Ravi; Nikhil Varma, P.; Jagadeesh, N.; Sandeep, J. V.; Cheepu, Muralimohan; Venkateswarlu, D.; Srinivas, B.

    2018-03-01

    Among the different renewable energy resources, solar energy is widely used due to its quantitative intensity factor. Solar air heater is cheap, simple in design and has got wide range of applications. A modest solar air heater has a lower in heat transfer and thermal performance as it has heat transfer coefficient lower in between coated absorber plate and the carrier fluid. This low thermal performance can be reduced to a greater extent by introducing the artificially created roughness over the absorber plate of the solar heater. In the present study, the combination of various geometries and roughness’s on the absorber plate are reported. Methods have been developed and implemented in order to improve the rate of the heat transfer. A comparison is drawn among different geometries to select the most effective absorber plate roughness. For flow analysis k-ω SST model was used and the constant heat flux was taken as 1100 W/m2. The Reynolds number is varied in a range from 3000 to 20000. The variation of different parameters temperature, Nusselt number, turbulence kinetic energy and heat transfer coefficient with Reynolds number were examined and discussed.

  3. Performance Study of Photovoltaic-Thermal (Pv/T) Solar Collector with ·-Grooved Absorber Plate

    International Nuclear Information System (INIS)

    Mohd Yusof Othman; Hafidz Ruslan; Kamaruzzaman Sopian; Jin, G.L.

    2009-01-01

    A hybrid photovoltaic-thermal solar collector has been designed, built and its performance has been studied. The advantage of the collector is that it can generate electricity and heat simultaneously. Photovoltaic module SHARP NE-80E2EA with maximum output power of 80 W was used to generate electricity. The module also acts as heat absorber of the collector. Single pass ·-groove collector made of aluminium sheet with 0.7 mm thickness has been used to collect heat generated. Study was conducted under a designed halogen lamps solar simulator with intensities set at 386 ± 8 Wm -2 and 817 ± 8 Wm -2 . The speed of air passing through the collector was set between (69.6 ± 2.2) x 10 -4 kg/s to (695.8 ± 2.2) x 10 -4 kg/s. The objective of the study is to compare the performance of PV/T collector with and without ·-groove absorber. The study found that the PV/T collector with ·-groove absorber plate has higher efficiency than the PV/T without ·-groove absorber. The electrical and thermal efficiencies are also increased when radiation intensity and speed of air increase. (author)

  4. ''In situ'' electronic testing method of a neutron detector performance

    International Nuclear Information System (INIS)

    Gonzalez, J.M.; Levai, F.

    1987-01-01

    The method allows detection of any important change in the electrical characteristics of a neutron sensor channel. It checks the response signal produced by an electronic detector circuit when a pulse generator is connected as input signal in the high voltage supply. The electronic circuit compares the detector capacitance value, previously measured, against a reference value, which is adjusted in a window type comparator electronic circuit to detect any important degrading condition of the capacitance value in a detector-cable system. The ''in-situ'' electronic testing method of neutron detector performance has been verified in a laboratory atmosphere to be a potential method to detect any significant change in the capacitance value of a nuclear sensor and its connecting cable, also checking: detector disconnections, cable disconnections, length changes of the connecting cable, electric short-opened circuits in the sensor channel, and any electrical trouble in the detector-connector-cable system. The experimental practices were carried out by simulation of several electric changes in a nuclear sensor-cable system from a linear D.C. channel which measures reactor power during nuclear reactor operation. It was made at the Training Reactor Electronic Laboratory. The results and conclusions obtained at the Laboratory were proved, satisfactorily, in the Electronic Instrumentation of Budapest Technical University Training Reactor, Hungary

  5. Year-round performance of a modified single-basin solar still with mica plate as a suspended absorber

    Energy Technology Data Exchange (ETDEWEB)

    El-Sebaii, A.A.; Aboul-Enein, S.; Ramadan, M.R.I.; El-Bialy, E. [Tanta University (Egypt). Faculty of Science

    2000-01-01

    In a previous study, a single-basin solar still with suspended absorber (SBSSBA) made from aluminium was constructed and investigated experimentally and theoretically. It was found that the daily productivity of the still was about 20% higher than that of the conventional single-basin solar still (SBSS). In this paper, the effect of thermal conductivity of the suspended absorber on the daily productivity of the still is investigated experimentally using aluminium, copper, stainless steel and mica plates as suspended absorbers. The results obtained are compared with those obtained for the SBSS tested under the same climatic conditions of Tanta (lat. 30{sup o} 47' N). The results indicate that it is advisable to use suspended plates made from insulating materials, such as mica, plastic, glass, etc. The daily productivity of the modified still with mica is found to be 42% higher than that of SBSS. Further, the effect of thickness of the suspended absorber on the productivity as well as the year-round performances of both SBSSBA and SBSS for the year 1996 are studied by computer simulation. There is good evidence that the productivity of SBSSBA is less dependent on the thickness of the suspended plate. The annual average productivities of the modified still with mica are found to be 23 and 15.8% higher than those of the conventional still when the basin water masses are 80 and 40 kg, respectively. This indicates that the suspended plate becomes more effective at higher masses of basin water. (author)

  6. Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei, E-mail: huj1@ornl.gov [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Road, PO Box 2008, MS-6172, Oak Ridge, TN 37831-6172 (United States); Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T. [Nuclear Engineering and Nonproliferation Division, Los Alamos National Laboratory (United States)

    2013-11-21

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the {sup 239}Pu and {sup 235}U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of {sup 239}Pu and {sup 235}U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that {sup 235}U content has a major impact on the SINRD signal in addition to the {sup 239}Pu content. Plutonium-241 and {sup 241}Am are the two main nuclides responsible for the variation in the count

  7. Performance of super-absorbent polymer as an internal curing agent for self-compacting concrete

    Directory of Open Access Journals (Sweden)

    Al-Hubboubi Suhair

    2018-01-01

    Full Text Available Internal curing agent by using super-absorbent polymer was present in this study, its effect on the properties of self-compacting concrete was evaluated .The SAP content in the concrete mix was 0.5 % by weight of cement. Three procedures for curing were adopted; curing in water, curing in water and air and curing in polyethylene sealed bags. Fresh concrete tests conducted to assess the self-compactability of the produced concrete. Moreover, compressive and splitting strength tests were carried out. The testing program had been extended to the age of 90 days.The use of super-absorbent polymer did not affect the fresh state characteristics of the studied SCC and achieved an increase in both compressive and tensile strengths as compared to the reference concrete mix.

  8. Performance analysis of solar air heater with jet impingement on corrugated absorber plate

    Directory of Open Access Journals (Sweden)

    Alsanossi M. Aboghrara

    2017-09-01

    Full Text Available This paper deals with the experimental investigation outlet temperature and efficiency, of Solar Air heater (SAH. The experimental test set up designed and fabricated to study the effect of jet impingement on the corrugated absorber plate, through circular jets in a duct flow of solar air heater, and compared with conventional solar air heater on flat plat absorber. Under effect of mass flow rate (ṁ of air and solar radiation on outlet air temperature, and efficiency, are analyzed. Results show the flow jet impingement on corrugated plat absorber is a strong function of heat transfer enhancement. The present investigation concludes that the mass flow rate of air substantially influences the heat transfer on solar air heaters. And the thermal efficiency of proposed design duct is observed almost 14% more as compare to the smooth duct. At solar radiation 500–1000 (W/M2, 308 K ambient temperature and 0.01–0.03 (Kg/S mass flow rate

  9. Imperfectly geometric shapes of nanograting structures as solar absorbers with superior performance for solar cells.

    Science.gov (United States)

    Nguyen-Huu, Nghia; Cada, Michael; Pištora, Jaromír

    2014-03-10

    The expectation of perfectly geometric shapes of subwavelength grating (SWG) structures such as smoothness of sidewalls and sharp corners and nonexistence of grating defects is not realistic due to micro/nanofabrication processes. This work numerically investigates optical properties of an optimal solar absorber comprising a single-layered silicon (Si) SWG deposited on a finite Si substrate, with a careful consideration given to effects of various types of its imperfect geometry. The absorptance spectra of the solar absorber with different geometric shapes, namely, the grating with attached nanometer-sized features at the top and bottom of sidewalls and periodic defects within four and ten grating periods are investigated comprehensively. It is found that the grating with attached features at the bottom absorbs more energy than both the one at the top and the perfect grating. In addition, it is shown that the grating with defects in each fourth period exhibits the highest average absorptance (91%) compared with that of the grating having defects in each tenth period (89%), the grating with attached features (89%), and the perfect one (86%). Moreover, the results indicate that the absorptance spectrum of the imperfect structures is insensitive to angles of incidence. Furthermore, the absorptance enhancement is clearly demonstrated by computing magnetic field, energy density, and Poynting vector distributions. The results presented in this study prove that imperfect geometries of the nanograting structure display a higher absorptance than the perfect one, and provide such a practical guideline for nanofabrication capabilities necessary to be considered by structure designers.

  10. Shock absorbing structure

    International Nuclear Information System (INIS)

    Kojima, Naoki; Matsushita, Kazuo.

    1992-01-01

    Small pieces of shock absorbers are filled in a space of a shock absorbing vessel which is divided into a plurality of sections by partitioning members. These sections function to prevent excess deformation or replacement of the fillers upon occurrence of falling accident. Since the shock absorbing small pieces in the shock absorbing vessel are filled irregularly, shock absorbing characteristics such as compression strength is not varied depending on the direction, but they exhibit excellent shock absorbing performance. They surely absorb shocks exerted on a transportation vessel upon falling or the like. If existing artificial fillers such as pole rings made of metal or ceramic and cut pieces such as alumium extrusion molding products are used as the shock absorbing pieces, they have excellent fire-proofness and cold resistance since the small pieces are inflammable and do not contain water. (T.M.)

  11. Enhanced performance of VOx-based bolometer using patterned gold black absorber

    Science.gov (United States)

    Smith, Evan M.; Panjwani, Deep; Ginn, James; Warren, Andrew; Long, Christopher; Figuieredo, Pedro; Smith, Christian; Perlstein, Joshua; Walter, Nick; Hirschmugl, Carol; Peale, Robert E.; Shelton, David J.

    2015-06-01

    Patterned highly absorbing gold black film has been selectively deposited on the active surfaces of a vanadium-oxide-based infrared bolometer array. Patterning by metal lift-off relies on protection of the fragile gold black with an evaporated oxide, which preserves gold black's near unity absorption. This patterned gold black also survives the dry-etch removal of the sacrificial polyimide used to fabricate the air-bridge bolometers. Infrared responsivity is substantially improved by the gold black coating without significantly increasing noise. The increase in the time constant caused by the additional mass of gold black is a modest 14%.

  12. Dataset demonstrating the modeling of a high performance Cu(In,GaSe2 absorber based thin film photovoltaic cell

    Directory of Open Access Journals (Sweden)

    Md. Asaduzzaman

    2017-04-01

    Full Text Available The physical data of the semiconductor materials used in the design of a CIGS absorber based thin film photovoltaic cell have been presented in this data article. Besides, the values of the contact parameter and operating conditions of the cell have been reported. Furthermore, by conducting the simulation with data corresponding to the device structure: soda-lime glass (SLG substrate/Mo back-contact/CIGS absorber/CdS buffer/intrinsic ZnO/Al-doped ZnO window/Al-grid front-contact, the solar cell performance parameters such as open circuit voltage (Voc, short circuit current density Jsc, fill factor (FF, efficiency (η, and collection efficiency ηc have been analyzed.

  13. Pinning down high-performance Cu-chalcogenides as thin-film solar cell absorbers: A successive screening approach

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yubo; Zhang, Wenqing, E-mail: wqzhang@mail.sic.ac.cn, E-mail: pzhang3@buffalo.edu [Materials Genome Institute and Department of Physics, Shanghai University, Shanghai 200444 (China); State Key Laboratory of High Performance Ceramics and Superfine Microstructures, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); Wang, Youwei; Zhang, Jiawei; Xi, Lili [State Key Laboratory of High Performance Ceramics and Superfine Microstructures, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); Zhang, Peihong, E-mail: wqzhang@mail.sic.ac.cn, E-mail: pzhang3@buffalo.edu [Materials Genome Institute and Department of Physics, Shanghai University, Shanghai 200444 (China); Department of Physics, University at Buffalo, SUNY, Buffalo, New York 14260 (United States)

    2016-05-21

    Photovoltaic performances of Cu-chalcogenides solar cells are strongly correlated with the absorber fundamental properties such as optimal bandgap, desired band alignment with window material, and high photon absorption ability. According to these criteria, we carry out a successive screening for 90 Cu-chalcogenides using efficient theoretical approaches. Besides the well-recognized CuInSe{sub 2} and Cu{sub 2}ZnSnSe{sub 4} materials, several novel candidates are identified to have optimal bandgaps of around 1.0–1.5 eV, spike-like band alignments with CdS window layer, sharp photon absorption edges, and high absorption coefficients. These new systems have great potential to be superior absorbers for photovolatic applications if their carrrier transport and defect properties are properly optimized.

  14. Performance of compact TES arrays with integrated high-fill-fraction X-ray absorbers

    International Nuclear Information System (INIS)

    Lindeman, Mark A.; Bandler, Simon; Brekosky, Regis P.; Chervenak, James A.; Figueroa-Feliciano, Enectali; Finkbeiner, Fred M.; Kelley, Richard L.; Saab, Tarek; Stahle, Caroline K.; Talley, D.J.

    2004-01-01

    We have recently produced and tested two-dimensional arrays of Mo/Au transition-edge-sensor (TES) calorimeters with Bi/Cu absorbers. The arrays represent a significant step towards meeting the specifications of NASA's Constellation-X mission. The calorimeters are compactly spaced within 5x5 arrays of 250 μm square pixels necessary for an angular resolution of 5 arcsec. Lithographically produced absorbers hang over the substrate and wiring between the TESs for high filling fraction and high quantum efficiency. We designed the calorimeters with heat capacities and thermal couplings such that X-rays produce pulses with fall times of approximately 300 μs to allow relatively high count rates with low dead time. We read out up to four of the pixels simultaneously. The arrays demonstrated very good energy resolution (5 eV at 1.5 keV and 7 eV at 6 keV) and little crosstalk between neighboring pixels

  15. Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

    Science.gov (United States)

    Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

    2003-10-01

    A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

  16. 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)

  17. Compilation of Existing Neutron Screen Technology

    Directory of Open Access Journals (Sweden)

    N. Chrysanthopoulou

    2014-01-01

    Full Text Available The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

  18. 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

  19. Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

    Science.gov (United States)

    Übeyli, Mustafa; Tel, Eyyüp

    2003-06-01

    Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

  20. Preparation of hollow microspheres of Ce{sup 3+} doped NiCo ferrite with high microwave absorbing performance

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Hong-zhen, E-mail: duanhz2000@163.com; Zhou, Fang-ling; Cheng, Xia; Chen, Guo-hong; Li, Qiao-ling

    2017-02-15

    Hollow microspheres of Ce{sup 3+} doped NiCo-ferrites were synthesized by template-based-deposition and surface reaction method with carbon sphere as the template. The phase structure, morphology, magnetic properties and wave absorbing properties of the sample were characterized by X-ray powder diffraction(XRD), Scanning electronic microscopy(SEM), Vibration sample magnetometer (VSM) and a network vector analyzer (NVA), respectively. The results indicated that the particle size of the carbon sphere sample prepared by hydrothermal method was about 0.5 µm and the particle size of the Ni{sub 0.5}Co{sub 0.5}Fe{sub 2}O{sub 4} sample prepared by template-based method was about 300 nm. The influence of the amount of rare earth element on the magnetic and absorbing properties of sample was studied. The saturation magnetization and coercivity decreased gradually with the increase of the content of Ce. When the content of Ce was 0.02, the maximal saturation magnetization value and coercivity was 75.72 emu• g{sup −1} and 789.88 Oe, respectively. The associated ferrite hollow spheres have good absorbing performance, and the return loss value was −18.8 dB at 5500 MHz. - Highlights: • Hollow microspheres of Ce{sup 3+} doped NiCo-ferrites were synthesized by template-based-deposition and surface reaction method. • The influence of rare earth Ce{sup 3+} on the magnetic and absorbing properties of sample was studied. • When the content of Ce was 0.02, the maximal saturation magnetization value and coercivity was 75.72 emu• g{sup −1} and 789.88 Oe, respectively.

  1. Neutron radiography of heated high-performance mortar

    Directory of Open Access Journals (Sweden)

    Weber B.

    2013-09-01

    Full Text Available Neutron radiography was applied to investigate the water distribution in mortar samples heated from one side to 600 °C. In mortar, aggregates and anhydrous cement are almost transparent to neutrons, while hydration products and water-filled capillary pores bear the largest attenuation. The evolution of the moisture profile shows a sharp dehydration front and accumulation of water due to condensation of water vapor behind this front.

  2. Neutronic design and performance analysis of Korean ITER TBM by Monte Carlo method

    International Nuclear Information System (INIS)

    Kim, Chang Hyo; Han, Beom Seok; Park, Ho Jin

    2006-01-01

    The objective of this project is to develop a neutronic design of the Korean TBM(Test Blanket Module) which will be installed in ITER(International Thermonuclear Experimental Reactor). This project is intended to analyze a neutronic design and nuclear performances of the Korean ITER TBM through the transport calculation of MCCARD. In detail, we will conduct numerical experiments for developing the neutronic design of the Korean ITER TBM and improving the nuclear performances. The results of the numerical experiments produced in this project will be utilized for a design optimization of the Korean ITER TBM. In this project, we proposed the neutronic methodologies for analyzing the nuclear characteristics of the fusion blanket. In order to investigate the behavior of neutrons and photons in the fusion blanket, Monte Carlo transport calculation was conducted with MCCARD. In addition, to optimize the neutronic performances of the fusion blanket, we introduced the design concept using a graphite reflector and a Pb multiplier. Through various numerical experiments, it was verified that these design concepts can be utilized efficiently to improve neutronic performances and resolve many drawbacks. The graphite-reflected HCML blanket can provide the neutronic performances far better than the non-reflected blanket, and a slightly-enriched Li breeder can satisfy the tritium self-sufficiency. The HCSB blanket design concept with a graphite reflector and a Pb multiplier was proposed. According to results of the neutronic analyses, the graphite-reflected HCSB blanket with a Pb multiplier can provide the neutronic performances comparable with those of the conventional HCSB blanket

  3. High Performance Infrared Plasmonic Metamaterial Absorbers and Their Applications to Thin-film Sensing

    KAUST Repository

    Yue, Weisheng

    2016-04-07

    Plasmonic metamaterial absorbers (PMAs) have attracted considerable attention for developing various sensing devices. In this work, we design, fabricate and characterize PMAs of different geometrical shapes operating in mid-infrared frequencies, and explore the applications of the PMAs as sensor for thin films. The PMAs, consisting of metal-insulator-metal stacks with patterned gold nanostructured surfaces (resonators), demonstrated high absorption efficiency (87 to 98 %) of electromagnetic waves in the infrared regime. The position and efficiency of resonance absorption are dependent on the shape of the resonators. Furthermore, the resonance wavelength of PMAs was sensitive to the thin film coated on the surface of the PMAs, which was tested using aluminum oxide (Al2O3) as the film. With increase of the Al2O3 thickness, the position of resonance absorption shifted to longer wavelengths. The dependence of the resonant wavelength on thin film thickness makes PMAs a suitable candidate as a sensor for thin films. Using this sensing strategy, PMAs have potential as a new method for thin film detection and in situ monitoring of surface reactions. © 2016 Springer Science+Business Media New York

  4. High Performance Infrared Plasmonic Metamaterial Absorbers and Their Applications to Thin-film Sensing

    KAUST Repository

    Yue, Weisheng; Wang, Zhihong; Yang, Yang; Han, Jiaguang; Li, Jingqi; Guo, Zaibing; Tan, Hua; Zhang, Xixiang

    2016-01-01

    Plasmonic metamaterial absorbers (PMAs) have attracted considerable attention for developing various sensing devices. In this work, we design, fabricate and characterize PMAs of different geometrical shapes operating in mid-infrared frequencies, and explore the applications of the PMAs as sensor for thin films. The PMAs, consisting of metal-insulator-metal stacks with patterned gold nanostructured surfaces (resonators), demonstrated high absorption efficiency (87 to 98 %) of electromagnetic waves in the infrared regime. The position and efficiency of resonance absorption are dependent on the shape of the resonators. Furthermore, the resonance wavelength of PMAs was sensitive to the thin film coated on the surface of the PMAs, which was tested using aluminum oxide (Al2O3) as the film. With increase of the Al2O3 thickness, the position of resonance absorption shifted to longer wavelengths. The dependence of the resonant wavelength on thin film thickness makes PMAs a suitable candidate as a sensor for thin films. Using this sensing strategy, PMAs have potential as a new method for thin film detection and in situ monitoring of surface reactions. © 2016 Springer Science+Business Media New York

  5. Performance of a Tethered Point Wave-Energy Absorber in Regular and Irregular Waves

    KAUST Repository

    Bachynski, Erin E.; Young, Yin Lu; Yeung, Ronald W.

    2010-01-01

    The importance of the mooring system on the dynamic response of a point-absorber type ocean-wave energy converter (WEC) is investigated using a frequency-domain approach. In order to ensure the safety of WECs, careful consideration of the response and resonance frequencies in all motions must be evaluated, including the effects of the mooring system. In this study, a WEC floater with a closed, flat bottom is modeled as a rigid vertical cylinder tethered by elastic mooring lines. The WEC hydrodynamic added mass and damping are obtained using established potential-flow methods, with additional damping provided by the energy-extraction system. The results show that the response of the WEC, and the corresponding power takeoff, varies with the diameter-to-draft (D=T) ratio, mooring system stiffness, and mass distribution. For a given wave climate in Northern California, near San Francisco, the heave energy extraction is found to be best for a shallow WEC with a soft mooring system, compared to other systems that were examined. This result assumes a physical limit (cap) on the motion which is related to the significant wave height to draft ratio. Shallow draft designs, however, may experience excessive pitch motions and relatively larger viscous damping. In order to mitigate the pitch response, the pitch radius of gyration should be small and the center of mass should be low. Copyright © 2010 by ASME.

  6. Multidisk neutron velocity selectors

    International Nuclear Information System (INIS)

    Hammouda, B.

    1992-01-01

    Helical multidisk velocity selectors used for neutron scattering applications have been analyzed and tested experimentally. Design and performance considerations are discussed along with simple explanation of the basic concept. A simple progression is used for the inter-disk spacing in the 'Rosta' design. Ray tracing computer investigations are presented in order to assess the 'coverage' (how many absorbing layers are stacked along the path of 'wrong' wavelength neutrons) and the relative number of neutrons absorbed in each disk (and therefore the relative amount of gamma radiation emitted from each disk). We discuss whether a multidisk velocity selector can be operated in the 'reverse' configuration (i.e. the selector is turned by 180 0 around a vertical axis with the rotor spun in the reverse direction). Experimental tests and calibration of a multidisk selector are reported together with evidence that a multidisk selector can be operated in the 'reverse' configuration. (orig.)

  7. 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

  8. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    International Nuclear Information System (INIS)

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

  9. A comprehensive simulation model of the performance of photochromic films in absorbance-modulation-optical-lithography

    Directory of Open Access Journals (Sweden)

    Apratim Majumder

    2016-03-01

    Full Text Available Optical lithography is the most prevalent method of fabricating micro-and nano-scale structures in the semiconductor industry due to the fact that patterning using photons is fast, accurate and provides high throughput. However, the resolution of this technique is inherently limited by the physical phenomenon of diffraction. Absorbance-Modulation-Optical Lithography (AMOL, a recently developed technique has been successfully demonstrated to be able to circumvent this diffraction limit. AMOL employs a dual-wavelength exposure system in conjunction with spectrally selective reversible photo-transitions in thin films of photochromic molecules to achieve patterning of features with sizes beyond the far-field diffraction limit. We have developed a finite-element-method based full-electromagnetic-wave solution model that simulates the photo-chemical processes that occur within the thin film of the photochromic molecules under illumination by the exposure and confining wavelengths in AMOL. This model allows us to understand how the material characteristics influence the confinement to sub-diffraction dimensions, of the transmitted point spread function (PSF of the exposure wavelength inside the recording medium. The model reported here provides the most comprehensive analysis of the AMOL process to-date, and the results show that the most important factors that govern the process, are the polarization of the two beams, the ratio of the intensities of the two wavelengths, the relative absorption coefficients and the concentration of the photochromic species, the thickness of the photochromic layer and the quantum yields of the photoreactions at the two wavelengths. The aim of this work is to elucidate the requirements of AMOL in successfully circumventing the far-field diffraction limit.

  10. A comprehensive simulation model of the performance of photochromic films in absorbance-modulation-optical-lithography

    Energy Technology Data Exchange (ETDEWEB)

    Majumder, Apratim; Helms, Phillip L.; Menon, Rajesh, E-mail: rmenon@eng.utah.edu [Department of Electrical and Computer Engineering, University of Utah, Salt Lake City, Utah 84112 (United States); Andrew, Trisha L. [Department of Chemistry, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States)

    2016-03-15

    Optical lithography is the most prevalent method of fabricating micro-and nano-scale structures in the semiconductor industry due to the fact that patterning using photons is fast, accurate and provides high throughput. However, the resolution of this technique is inherently limited by the physical phenomenon of diffraction. Absorbance-Modulation-Optical Lithography (AMOL), a recently developed technique has been successfully demonstrated to be able to circumvent this diffraction limit. AMOL employs a dual-wavelength exposure system in conjunction with spectrally selective reversible photo-transitions in thin films of photochromic molecules to achieve patterning of features with sizes beyond the far-field diffraction limit. We have developed a finite-element-method based full-electromagnetic-wave solution model that simulates the photo-chemical processes that occur within the thin film of the photochromic molecules under illumination by the exposure and confining wavelengths in AMOL. This model allows us to understand how the material characteristics influence the confinement to sub-diffraction dimensions, of the transmitted point spread function (PSF) of the exposure wavelength inside the recording medium. The model reported here provides the most comprehensive analysis of the AMOL process to-date, and the results show that the most important factors that govern the process, are the polarization of the two beams, the ratio of the intensities of the two wavelengths, the relative absorption coefficients and the concentration of the photochromic species, the thickness of the photochromic layer and the quantum yields of the photoreactions at the two wavelengths. The aim of this work is to elucidate the requirements of AMOL in successfully circumventing the far-field diffraction limit.

  11. Discrimination of various contributions to the absorbed dose in BNCT: Fricke-gel imaging and intercomparison with other experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Gambarini, G. E-mail: grazia.gambarini@mi.infn.it; Agosteo, S.; Marchesi, P.; Nava, E.; Palazzi, P.; Pecci, A.; Rosi, G.; Tinti, R

    2000-11-15

    A method is described for the 3D measurements of absorbed dose in a ferrous sulphate gel phantom, exposed in the thermal column of a nuclear reactor. The method, studied for Boron Neutron Capture Therapy (BNCT) purposes, allows absorbed dose imaging and profiling, with the separation of different contributions coming from different secondary radiations, generated from thermal neutrons. In fact, the biological effectiveness of the different radiations is different. Tests with conventional dosimeters were performed too.

  12. New spallation neutron sources, their performance and applications

    International Nuclear Information System (INIS)

    1985-01-01

    Pulsed spallation sources now operating in the world are at the KEK Laboratory in Japan (the KENS source), at Los Alamos National Laboratory (WNR) and at Argonne National Laboratory (IPNS), both the latter being in the US. The Intense Pulsed Neutron Source (IPNS) is currently the world's most intense source with a peak neutron flux of 4 x 10 14 n cm -2 s -1 at a repetition rate of 30 Hz, and globally producing approx. 1.5 x 10 15 n/sec. Present pulsed sources are still relatively weak compared to their potential. In 1985 the Rutherford Spallation Neutron Source will come on line, and eventually be approx. 30 more intense than the present IPNS. Later, in 1986 the WNR/PSR option at Los Alamos will make that facility of comparable intensity, while a subcritical fission booster at IPNS will keep IPNS competitive. These new sources will expand the applications of pulsed neutrons but are still based on accelerators built for other scientific purposes, usually nuclear or high-energy physics. Accelerator physicists are now designing machines expressly for spallation neutron research, and the proton currents attainable appear in the milliamps. (IPNS now runs at 0.5 GeV and 14 μA). Such design teams are at the KFA Laboratory Julich, Argonne National Laboratory and KEK. Characteristics, particularly the different time structure of the pulses, of these new sources will be discussed. Such machines will be expensive and require national, if not international, collaboration across a wide spectrum of scientific disciplines. The new opportunities for neutron research will, of course, be dramatic with these new sources

  13. Neutron-scattering study of the vibrational behavior of trehalose aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Branca, C.; Magazu, S.; Migliardo, F.; Romeo, G.; Villari, V.; Wanderlingh, U. [Dipartimento di Fisica and INFM, Universita' di Messina, PO Box 55, 98166 Messina (Italy); Colognesi, D. [DRAL-ISIS,Chilton, Oxford OX1 3PU (United Kingdom)

    2002-07-01

    Neutron spectra for hydrated trehalose samples have been obtained by using the time-of-flight spectrometer TOSCA at the ISIS Pulse Neutron Facility (Rutherford Appleton Laboratory, Chilton, UK). Neutron spectra have been compared to the absorbance spectra obtained by Fourier-transform infrared spectroscopy. Finally, a comparison with findings obtained by density functional theory has been performed. (orig.)

  14. Developments of high-performance moderator vessel for JRR-3 cold neutron source

    International Nuclear Information System (INIS)

    Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

    2015-05-01

    The cold neutron source (CNS) facility converts thermal neutrons into cold neutrons to moderate neutrons with liquid hydrogen. The cold neutron beam at Japan Research Reactor No. 3 (JRR-3) is led to the beam experimental devices in the beam hall through neutron guide tubes. High intensities of the cold neutron beam are always demanded for increasing the experimental effectiveness and accuracy. In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3. (author)

  15. 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)

  16. Modelling the performance parameters of a horizontal falling film absorber with aqueous (lithium, potassium, sodium) nitrate solution using artificial neural networks

    International Nuclear Information System (INIS)

    Álvarez, María E.; Hernández, José A.; Bourouis, Mahmoud

    2016-01-01

    An ANN (artificial neural network) model was developed to determine the efficiency parameters of a horizontal falling film absorber at operating conditions of interest for absorption cooling systems. The aqueous nitrate solution LiNO_3 + KNO_3 + NaNO_3 with salt mass percentages of 53%, 28% and 19%, respectively, was used as a working fluid. The authors created the ANN from the database they had compiled with the results of experiments that they had performed in a set-up designed and built for this purpose. The ANN structure consisted of 6 input variables: inlet solution and cooling water temperatures, cooling water and solution mass flow rates, absorber pressure and inlet solution concentration; 4 output variables which facilitated the assessment of the performance of the absorber: heat and mass transfer coefficients, absorption mass flux and the degree of subcooling of the solution leaving the absorber. The hidden layer contained 9 neurons which were determined by training and test procedures. The results showed that the deviation between the experimental data and the estimated values was well adjusted. This indicated that the ANN model was an effective tool for predicting the efficiency parameters of the absorber. The solution flow rate was also observed to be the most significant operating variable which affected the performance of the absorber. - Highlights: • An ANN was developed to predict the efficiency parameters of a falling film absorber. • The ANN was created using a database corresponding to a triple-effect absorption chiller. • The ANN predicts the efficiency parameters of falling film absorbers with r"2 > 0.95. • The solution flow rate is the variable that most affects the performance of the absorber.

  17. Thermo-hydraulic performance of solar air heater having multiple v-shaped rib roughness on absorber plates

    Directory of Open Access Journals (Sweden)

    Dhananjay Kumar

    2018-03-01

    Full Text Available This paper presents the performance analysis of the effect of geometrical parameters having multiple v-shaped rib roughness on the airflow side of the absorber plates. Mathematical approach and solution procedure for the analysis of such a solar air heater has been developed theoretically and MATLAB code generated for the solution of the mathematical equations. The effect of parameters such as flow Reynolds number and Relative roughness height on the thermohydraulic performance have been examined and compared with the conventional flat plate solar air heater. A substantial improvement in thermal efficiency of roughened solar air heater as compared to smooth one due to appreciable enhancement in heat transfer coefficient. The enhancement in heat transfer coefficient is also accompanied by a considerable enhancement in pumping power requirement due to the increase in friction factor.

  18. Using FLUKA to Study Concrete Square Shield Performance in Attenuation of Neutron Radiation Produced by APF Plasma Focus Neutron Source

    Science.gov (United States)

    Nemati, M. J.; Habibi, M.; Amrollahi, R.

    2013-04-01

    In 2010, representatives from the Nuclear Engineering and physics Department of Amirkabir University of Technology (AUT) requested development of a project with the objective of determining the performance of a concrete shield for their Plasma Focus as neutron source. The project team in Laboratory of Nuclear Engineering and physics department of Amirkabir University of Technology choose some shape of shield to study on their performance with Monte Carlo code. In the present work, the capability of Monte Carlo code FLUKA will be explored to model the APF Plasma Focus, and investigating the neutron fluence on the square concrete shield in each region of problem. The physical models embedded in FLUKA are mentioned, as well as examples of benchmarking against future experimental data. As a result of this study suitable thickness of concrete for shielding APF will be considered.

  19. A comparative performance study of a photovoltaic concentrator system with discrete mirror and continuos profile for two different absorber shapes

    Energy Technology Data Exchange (ETDEWEB)

    H, Saiful; Rezau, K.M [University of Dhaka, Dhaka (Bangladesh)

    2000-07-01

    Profiles of parabolic concentrators of discrete mirror and continuos surface mirror have been designed for combined electrical thermal photovoltaic systems. In the design the changes of concentration ratio, effect of reflection, angle of incidence over the absorber have been taken into account for maximum energy collection. The performances of the system are studied for solar cells of modified grid finger for illuminations from 1-10 sun. The local concentration ratio (LCR) distribution over the absorbers for both the concentrator, the optical efficiency, thermal efficiency, electrical and thermal power output and overall efficiency have been evaluated for different values of beam radiation concentration ratio and focal distance. [Spanish] Se han disenado perfiles de concentradores parabolicos de espejo discreto y de superficie continua para sistemas fotovoltaicos combinados electricos y termicos. En el diseno los cambios de la proporcion de concentracion, del efecto de la reflexion, del angulo de incidencia sobre el observador se han tenido en cuenta para una maxima recoleccion de energia. Los rendimientos del sistema se han estudiado para celdas solares de parrilla modificada para iluminaciones solares de 1-10. Han sido evaluados para diferentes valores de la proporcion de la concentracion de la radiacion en el rayo y la distancia focal la proporcion de concentracion local (LCR) de la distribucion en los absorbedores, para el concentrador la eficiencia optica, la eficiencia termica, electrica, la produccion de energia termica y electrica y la eficiencia total.

  20. Hydrothermal synthesis of magnetic Fe{sub 3}O{sub 4}/graphene composites with good electromagnetic microwave absorbing performances

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Lingyu; Zeng, Xiaojun; Li, Xiaopan; Yang, B., E-mail: byang@buaa.edu.cn; Yu, Ronghai, E-mail: rhyu@buaa.edu.cn

    2017-03-15

    The Fe{sub 3}O{sub 4} sub-microspheres have been embedded uniformly into the reduced graphene oxide (rGO) to form a new-type Fe{sub 3}O{sub 4}/rGO composites through a one-pot solvothermal method. The dielectric properties for these magnetic Fe{sub 3}O{sub 4}/rGO composites can be greatly tuned by their different rGO additions. A good impedance matching from the balanced dielectric and magnetic loss is achieved in the Fe{sub 3}O{sub 4}/rGO composites with 4 wt% rGO addition, which dominates their excellent microwave absorbing performances including the minimum reflection loss (RL) value of −45 dB at a frequency of 8.96 GHz with a sample thickness of 3.5 mm and an effective absorption bandwidth of 3.2 GHz (below −10 dB) superior to those of the most magnetic materials and carbon-based composites. The controlled Fe{sub 3}O{sub 4}/rGO composite structure also exhibits high chemical stability and low density, which shows great potential application in high-performance electromagnetic microwave-absorbing materials. - Highlights: • Magnetic Fe{sub 3}O{sub 4}/rGO composites are fabricated by a facile solvothermal method. • The dielectric properties for the Fe{sub 3}O{sub 4}/rGO composites can be tuned. • The Fe{sub 3}O{sub 4}/rGO composites exhibits high chemical stability and low density. • Excellent microwave absorption performances for the composites are obtained.

  1. Recent performance of the Intense Pulsed Neutron Source accelerator system

    International Nuclear Information System (INIS)

    Potts, C.; Brumwell, F.; Rauchas, A.; Stipp, V.; Volk, G.; Donley, L.

    1987-03-01

    The Intense Pulsed Neutron Source (IPNS) accelerator system has now been in operation as part of a national user program for over five years. During that period steady progress has been made in both beam intensity and reliability. Almost 1.8 billion pulses totaling 4 x 10 21 protons have now been delivered to the spallation neutron target. Recent weekly average currents have reached 15 μA (3.2 x 10 12 protons per pulse, 30 pulses per second) and short-term peaks of almost 17 μA have been reached. In fact, the average current for the last two years is up 31% over the average for the first three years of operation

  2. Performance of the intense pulsed neutron source accelerator system

    International Nuclear Information System (INIS)

    Potts, C.; Brumwell, F.; Rauchas, A.; Stipp, V.; Volk, G.

    1983-01-01

    The Intense Pulsed Neutron Source (IPNS) facility has now been operating in a routine way for outside users since November 1, 1981. From that date through December of 1982, the accelerator system was scheduled for neutron science for 4500 hours. During this time the accelerator achieved its short-term goals by delivering about 380,000,000 pulses of beam totaling over 6 x 10 20 protons. The changes in equipment and operating practices that evolved during this period of intense running are described. The intensity related instability threshold was increased by a factor of two and the accelerator beam current has been ion source limited. Plans to increase the accelerator intensity are also described. Initial operating results with a new H - ion source are discussed

  3. Performance of novel moderator for pulsed neutron diffraction

    International Nuclear Information System (INIS)

    Mayer, R.E.; Granada, J.R.; Dawidowski, J.; Gillette, V.H.

    1991-01-01

    Measurements of neutron pulse time-width and intensity have been carried out on grids of small moderators placed side by side and decoupled by cadmium strips. This moderator concept had been introduced at ICANS-10. The present measurements explore greater moderator thicknesses than those previously attained, yielding information on thickness optimization, while confirming the previous results on resolution which make this moderator a favourable choice in front of the conventional sandwich set-up. (author)

  4. Neutronic performance of Indian LLCB TBM set conceptual design in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Swami, H.L., E-mail: hswami@ipr.res.in; Shaw, A.K.; Mistry, A.N.; Danani, C.

    2016-12-15

    Highlights: • Neutronic analyses of conceptual design of LLCB test blanket module in ITER have been performed. • The estimated total tritium production rate in the LLCB TBM is 1.66E + 17 tritons/s. • Total heat deposited in the LLCB TBM is 0.46 MW and highest power density at TBM first wall is 5.2 Watt/cc. • The estimation shows the maximum DPA 2.72 at TBM FW. - Abstract: Tritium breeding blanket testing program in ITER is an important milestone towards the development of the fusion reactors. ITER organization is providing an opportunity to the partner countries to test their breeding blanket concepts. A mock-up of Indian Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket known as LLCB Test Blanket Module (TBM) will be tested in ITER equatorial port no. 2. LLCB blanket consists of lead lithium (PbLi) as a neutron multiplier & tritium breeder, ceramic breeder (Li{sub 2}TiO{sub 3}) as a tritium breeder and India specific Reduced Activation Ferretic Martinic Steel (IN-RAFMS) as a structural material. A stainless steel block which is cooled by water, called as shield block, is attached with TBM to provide neutron shield to ITER TBM port. A comprehensive neutronic performance evaluation is required for the design of the LLCB TBM set (TBM + shield block) and associated ancillary systems in ITER. The neutronic performance of the conceptual design of TBM set in ITER has been carried out and reported here. In order to carry out the neutronic performance evaluation, the neutronic models of the LLCB TBM set along with TBM frame have been constructed and inserted in the equatorial port of ITER reference neutronic model C-lite. Neutronic responses such as tritium production rate, nuclear heating, neutron flux & spectra, gas production & DPA in the LLCB TBM set are calculated considering 500 MW fusion power & fluence level of 0.3 MWa/m{sup 2}. Radiation transport code MCNP6 and FENDL 2.1 nuclear cross-section data library are used to perform the neutronic

  5. Stability of the Hall sensors performance under neutron irradiation

    International Nuclear Information System (INIS)

    Duran, I.; Hron, M.; Stockel, J.; Viererbl, L.; Vsolak, R.; Cerva, V.; Bolshakova, I.; Holyaka, R.; Vayakis, G.

    2004-01-01

    A principally new diagnostic method must be developed for magnetic measurements in steady state regime of operation of fusion reactor. One of the options is the use of transducers based on Hall effect. The use of Hall sensors in ITER is presently limited by their questionable radiation and thermal stability. Issues of reliable operation in ITER like radiation and thermal environment are addressed in the paper. The results of irradiation tests of candidate Hall sensors in LVR-15 and IBR-2 experimental fission reactors are presented. Stable operation (deterioration of sensitivity below one percent) of the specially prepared sensors was demonstrated during irradiation by the total fluence of 3.10 16 n/cm 2 in IBR-2 reactor. Increasing the total neutron fluence up to 3.10 17 n/cm 2 resulted in deterioration of the best sensor's output still below 10% as demonstrated during irradiation in LVR-15 fission reactor. This level of neutron is already higher than the expected ITER life time neutron fluence for a sensor location just outside the ITER vessel. (authors)

  6. Performance of Polycrystalline Photovoltaic and Thermal Collector (PVT on Serpentine-Parallel Absorbers Design

    Directory of Open Access Journals (Sweden)

    Mustofa Mustofa

    2017-03-01

    Full Text Available This paper presents the performance of an unglazed polycrystalline photovoltaic-thermal PVT on 0.045 kg/s mass flow rate. PVT combine photovoltaic modules and solar thermal collectors, forming a single device that receive solar radiation and produces heat and electricity simultaneously. The collector figures out serpentine-parallel tubes that can prolong fluid heat conductivity from morning till afternoon. During testing, cell PV, inlet and outlet fluid temperaturs were recorded by thermocouple digital LM35 Arduino Mega 2560. Panel voltage and electric current were also noted in which they were connected to computer and presented each second data recorded. But, in this performance only shows in the certain significant time data. This because the electric current was only noted by multimeter device not the digital one. Based on these testing data, average cell efficieny was about 19%, while thermal efficiency of above 50% and correspondeng cell efficiency of 11%, respectively

  7. Neutron detection performance of silicon carbide and diamond detectors with incomplete charge collection properties

    Energy Technology Data Exchange (ETDEWEB)

    Hodgson, M., E-mail: michael.hodgson@becq.co.uk [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Lohstroh, A.; Sellin, P. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Thomas, D. [NPL, Teddington TW11 0LW (United Kingdom)

    2017-03-01

    The benefits of neutron detection and spectroscopy with carbon based, wide band gap, semiconductor detectors have previously been discussed within the literature. However, at the time of writing there are still limitations with these detectors related to availability, cost, size and perceived quality. This study demonstrates that lower quality materials—indicated by lower charge collection efficiency (CCE), poor resolution and polarisation effect—available at wafer scale and lower cost, can fulfil requirements for fast neutron detection and spectroscopy for fluxes over several orders of magnitude, where only coarse energy discrimination is required. In this study, a single crystal diamond detector (D-SC, with 100% CCE), a polycrystalline diamond (D-PC, with ≈4% CCE) and semi-insulating silicon carbide (SiC-SI, with ≈35% CCE) have been compared for alpha and fast neutron performance. All detectors demonstrated alpha induced polarisation effects in the form of a change of both energy peak position and count rate with irradiation time. Despite these operational issues the ability to detect fast neutrons and distinguish neutron energies was observed. This performance was demonstrated over a wide dynamic range (500–40,000 neutrons/s), with neutron induced polarisation being demonstrated in D-PC and SiC-SI at high fluxes.

  8. An easier method for performing a pancreaticojejunostomy for the soft pancreas using a fast-absorbable suture

    Institute of Scientific and Technical Information of China (English)

    Kenichi Hakamada; Shunji Narumi; Yoshikazu Toyoki; Masaki Nara; Kenosuke Ishido; Takuya Miura; Norihito; Mutsuo Sasaki

    2008-01-01

    AIM:To clarify the usefulness of a new method for performing a pancreaticojejunostomy by using a fastabsorbable suture material irradiated polyglactin 910,and a temporary stent tube for a narrow pancreatic duct with a soft pancreatic texture.METHODS:Among 63 consecutive patients with soft pancreas undergoing a pancreaticoduodenectomy from 2003 to 2006,35 patients were treated with a new reconstructive method.Briefly,after the pancreatic transaction,a stent tube was inserted into the lumen of the pancreatic duct and ligated with it by a fastabsorbable suture.Another tip of the stent tube was introduced into the intestinal lumen at the jejunal limb,where a purse-string suture was made by another fast-absorbable suture to roughly fix the tube.The pancreaticojejunostomy was completed by ligating two fast-absorbable sutures to approximate the ductal end and the jejunal mucosa,and by adding a rough anastomosis between the pancreatic parenchyma and the seromuscular layer of the jejunum.The initial surgical results with this method were retrospectively compared with those of the 28 patients treated with conventional duct-to-mucosa anastomosis.RESULTS:The incidences of postoperative morbidity including pancreatic fistula were comparable between the two groups (new;3%-17% vs conventional;7%-14% according to the definitions).There was no mortality and re-admission.Late complications were also rarely seen.CONCLUSION:A pancreaticojejunostomy using an irradiated polyglactin 910 suture material and a temporary stent is easy to perform and is feasible even in cases with a narrow pancreatic duct and a normal soft pancreas.

  9. Flux gain for a next-generation neutron reflectometer resulting from improved supermirror performance

    CERN Document Server

    Rehm, C

    2002-01-01

    Next-generation spallation neutron source facilities will offer instruments with unprecedented capabilities through simultaneous enhancement of source power and usage of advanced optical components. The Spallation Neutron Source (SNS), already under construction at Oak Ridge National Laboratory and scheduled to be completed by 2006, will provide greater than an order of magnitude more effective source flux than current state-of-the-art facilities, including the most advanced research reactors. An additional order of magnitude gain is expected through the use of new optical devices and instrumentation concepts. Many instrument designs require supermirror neutron guides with very high critical angles for total reflection. In this contribution, we will discuss how the performance of a modern neutron-scattering instrument depends on the efficiency of these supermirrors. We summarize current limitations of supermirror coatings and outline ideas for enhancing their performance, particularly for improving the reflec...

  10. Improving the FGD absorber and ESP performance at Iskenderun power plant

    Energy Technology Data Exchange (ETDEWEB)

    Atmaca, Guerkan [ISKEN Enerji Uretim A.S., Ceyhan-Adana (Turkey); Stratmann, Werner; Wortmann, Birgit [STEAG Energy Services GmbH, Essen (Germany)

    2012-07-01

    The Iskenderun power plant is located at the bay of Iskenderun in the Province of Adana in the south of Turkey. Two units with a total output of 1,210 MW are operated. The annual fuel - imported coal from Colombia and South Africa - consumption amounts to about 3.3 million tonnes tce. To meet the SO{sub x} and particulate limits values it was necessary to improve the performance of the FGD scrubber and the electrostatic precipitator (ESP). In the first step, the potential of a fluid flow optimisation of both the scrubber and the ESP was determined by simulating the 'as build' situation (reference cases) with a computational fluid dynamics (CFD) code. In the second step, several possible variants of component modifications and structures were analysed and evaluated. In a last step the most improving modifications were proposed. (orig.)

  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. Research performed at the ET-RR-1 reactor using the neutron scattering equipment

    International Nuclear Information System (INIS)

    Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.

    1990-02-01

    This report represents the results of studies and measurements, performed at the ET-RR-1 reactor, using the neutron scattering equipment supplied by the IAEA according to the technical assistance project EGY/1/11/10. The results of these studies, starting in 1980 and continuing to date, are discussed; the use of the equipment, both as a neutron monochromator and fixed scattering angle spectrometer, is also assessed. (author). 19 refs, 17 figs

  13. Using leg muscles as shock absorbers: theoretical predictions and experimental results of drop landing performance.

    Science.gov (United States)

    Minetti, A E; Ardigò, L P; Susta, D; Cotelli, F

    1998-12-01

    The use of muscles as power dissipators is investigated in this study, both from the modellistic and the experimental points of view. Theoretical predictions of the drop landing manoeuvre for a range of initial conditions have been obtained by accounting for the mechanical characteristics of knee extensor muscles, the limb geometry and assuming maximum neural activation. Resulting dynamics have been represented in the phase plane (vertical displacement versus speed) to better classify the damping performance. Predictions of safe landing in sedentary subjects were associated to dropping from a maximum (feet) height of 1.6-2.0 m (about 11 m on the moon). Athletes can extend up to 2.6-3.0 m, while for obese males (m = 100 kg, standard stature) the limit should reduce to 0.9-1.3 m. These results have been calculated by including in the model the estimated stiffness of the 'global elastic elements' acting below the squat position. Experimental landings from a height of 0.4, 0.7, 1.1 m (sedentary males (SM) and male (AM) and female (AF) athletes from the alpine ski national team) showed dynamics similar to the model predictions. While the peak power (for a drop height of about 0.7 m) was similar in SM and AF (AM shows a +40% increase, about 33 W/kg), AF stopped the downward movement after a time interval (0.219 +/- 0.030 s) from touch-down 20% significantly shorter than SM. Landing strategy and the effect of anatomical constraints are discussed in the paper.

  14. Performance of Hollow Fiber Membrane Gas-Liquid Contactors to Absorb CO2 Using Diethanolamine (Dea as a Solvent

    Directory of Open Access Journals (Sweden)

    Sutrasno Kartohardjono

    2010-10-01

    Full Text Available This study uses DEA solution to absorb CO2 from the gas flow through the hollow fiber membrane contactors. This study aims to evaluate the performance of hollow fiber membrane contactors to absorb CO2 gas using DEA solution as solvent through mass transfer and hydrodynamics studies. The use of DEA solution is to reduce the mass transfer resistance in the liquid phase, and on the other side, the large contact area of the membrane surface can cover the disadvantage of membrane contactors; additional mass transfer resistance in the membrane phase. During experiments, CO2 feed flows through the fiber lumens, while the 0.01 M DEA solution flows in the shell side of membrane contactors. Experimental results show that the mass transfer coefficients and fluxes of CO2 increase with an increase in both water and DEA solution flow rates. Increasing the amount of fibers in the contactors will decrease the mass transfer and fluxes at the same DEA solution flow rate. Mass transfer coefficients and CO2 fluxes using DEA solution can achieve 28,000 and 7.6 million times greater than using water as solvent, respectively. Hydrodynamics studies show that the liquid pressure drops in the contactors increase with increasing liquid flow rate and number of fibers in the contactors. The friction between water and the fibers in the contactor was more pronounced at lower velocities, and therefore, the value of the friction factor is also higher at lower velocities.

  15. Parametric Study on the Thermal Performance and Optimal Design Elements of Solar Air Heater Enhanced with Jet Impingement on a Corrugated Absorber Plate

    Directory of Open Access Journals (Sweden)

    Alsanossi M. Aboghrara

    2018-01-01

    Full Text Available Previous works revealed that cross-corrugated absorber plate design and jet impingement on a flat absorber plate resulted in a significant increase in the performance of a solar air heater (SAH. Involving these two designs into one continuous design to improve the SAH performance remains absent in the literature. This study aimed to evaluate the achieved enhancement on performance parameters of a SAH with jet impingement on a corrugated absorber plate. An energy balance model was developed to compare the performance parameters of the proposed SAH with the other two SAHs. At a clear sky day and a mass flow rate of 0.04 kg/s, the hourly results revealed that the max fluid outlet temperatures for the proposed SAH, jet-to-flat plate SAH, and cross-corrugated plate SAH are 321, 317, and 313 K, respectively; the max absorber plate temperatures are 323.5, 326.5, and 328 K, respectively; the maximum temperature differences between the absorber plate and fluid outlet are ~3, 9, and 15 K, respectively; the max efficiencies are 65.7, 64.8, and 60%, respectively. Statistical t-test results confirmed significant differences between the mean efficiency of the proposed SAH and SAH with jet-to-flat plate. Hence, the proposed design is considered superior in improving the performance parameters of SAH compared to other designs.

  16. The effects of using of mineral and organic toxin absorbents on broiler performance and internal organs weight in experimental aflatoxicosis

    Directory of Open Access Journals (Sweden)

    Behnam Heidarpour

    2016-04-01

    Full Text Available Introduction The occurrence of mycotoxins in foods and feeds is a problem of major concern in all over the world. Profitability of poultry production can be greatly affected due to the frequency of feed contamination and the detrimental effects of these toxins on the performance. Aflatoxins, a group of closely related and biologically active mycotoxins, are produced by strains of Aspergillus flavus and Aspergillus parasiticus. They commonly occur as natural contaminant of poultry feeds. Domestic animal species such as chickens, ducks, cattle and turkeys consuming sublethal doses of aflatoxins for several days developed a toxic syndrome in which liver damage was the most significant change. The biological effects of aflatoxins could be categorized into two groups, long term and short term effects. Long term effects included chronic toxicity, cancer, birth defects and genetic alterations. Aflatoxins affected all poultry species, although they generally take relatively high levels to cause mortality, low levels can be detrimental if continually fed. Material and Methods This study was conducted to determine the efficacy of mineral, organic toxin absorbents, humic acid and yeast cell wall on performance and internal organs weight of broilers in experimental aflatoxicosis. This study was conducted in a completely randomize design with 432 Ross-308 broilers with 9 treatments, 4 replicates and 12 broilers in each replicate. Treatments included diet without aflatoxin, 2: diet contaminated with aflatoxin, 3: diet contaminated with aflatoxin and supplemented with 0.20 Humic acid, 4: diet contaminated with aflatoxin and supplemented with 0.40 Humic acid, 5: diet contaminated with aflatoxin and supplemented with 0.60 Humic acid, 5: diet contaminated with aflatoxin and supplemented with 0.80 Humic acid, 6: diet contaminated with aflatoxin and supplemented with 0.80 Humic acid, 7: diet contaminated with aflatoxin and supplemented with 1.00 Humic acid, 8: diet

  17. Scintillating-Glass-Fiber neutron sensors, their application and performance for plutonium detection and monitoring

    International Nuclear Information System (INIS)

    Seymour, R.S.; Richardson, B.; Morichi, M.; Bliss, M.; Craig, R.A.; Sunberg, D.S.

    1998-01-01

    Most neutron detection sensors presently employ 3 He gas-filled detectors. Despite their excellent performance and widespread use, there are significant limitations to this technology. A significant alternative neutron sensor utilizing neutron-active material incorporated into a glass scintillator is presented that offers novel commercial sensors not possible or practical with gas tube technology. The scintillating optical fiber permits sensors with a multitude of sizes ranging from devices of a single fiber of 150μm to sensors with tens of thousands of fibers with areas as large as 5m 2 depending on the neutron flux to be measured. A second significant advantage is the use of high-speed electronics that allow a greater dynamic range, not possible with gas detectors. These sensors are flexible, conformable and less sensitive to vibration that optimizes the source-to-detector geometry and provides robust performance in field applications. The glass-fibers are sensitive to both gamma rays and neutrons. However the coincidence electronics are optimized for neutron to gamma ray discrimination allowing very sensitive measurements with a low false-alarm rate. Applications include SNM surveillance, material control and accountability (MC and A), safeguard inspections, Pu health physics / bioassay and environmental characterization. (author)

  18. Initial performance of the Cornell cold neutron beam

    International Nuclear Information System (INIS)

    Clark, D.D.; Spern, S.A.; Atwood, A.G.

    1997-01-01

    The cold source for a guided neutron beam has been installed in a Cornell TRIGA beamport and has successfully undergone thermal tests up to full power (normally 480 kW). Tests to date (8/1/96) include spectral and yield measurements at 10 kW with the first three meters of the 2-cm by 5-cm Ni-on-glass guide in place. A 110-cm 3 Al chamber, located 17 cm from the core, contains solid mesitylene and is cooled by conduction through a 269-cm long Cu rod connected to a cryorefrigerator outside the reactor shield. Distributions of flux per unit velocity have been measured at 10 kW by time-of-flight. Anticipated properties of the complete 13 m long beam at full power are discussed. (author)

  19. 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.

  20. 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

  1. 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.

  2. High performance and thermally stable tandem solar selective absorber coating for concentrated solar thermal power (CSP) application

    Science.gov (United States)

    Prasad, M. Shiva; Kumar, K. K. Phani; Atchuta, S. R.; Sobha, B.; Sakthivel, S.

    2018-05-01

    A novel tandem absorber system (Mn-Cu-Co-Ox-ZrO2/SiO2) developed on an austenitic stainless steel (SS-304) substrate to show an excellent optical performance (αsol: 0.96; ɛ: 0.23@500 °C). In order to achieve this durable tandem, we experimented with two antireflective layers such as ZrO2-SiO2 and nano SiO2 layer on top of Mn-Cu-Co-Ox-ZrO2 layer. We optimized the thickness of antireflective layers to get good tandem system in terms of solar absorptance and emittance. Field emission scanning electron microscopy (FESEM), UV-Vis-NIR and Fourier transform infrared spectroscopy (FTIR) were used to characterize the developed coatings. Finally, the Mn-Cu-Co-Ox-ZrO2/SiO2 exhibits high temperature resistance up to 800 °C, thus allow an increase in the operating temperature of CSP which may lead to high efficiency. We successfully developed a high temperature resistant tandem layer with easy manufacturability at low cost which is an attractive candidate for concentrated solar power generation (CSP).

  3. The performance of neutron spectrometers AR a long-pulse spallation source

    International Nuclear Information System (INIS)

    Pynn, R.; Daemen, L.L.

    1995-01-01

    At a recent workshop at Lawrence Berkeley National Laboratory members of the international neutron scattering community discussed the performance to be anticipated from neutron scattering instruments installed at a 1 MW long-pulse spallation source (LPSS). Although the report of this workshop is long, its principal conclusions can be easily summarised and almost as easily understood. This article presents such a synthesis for a 60 Hz LPSS with 1 msec proton pulses. We discuss some of the limitations of the workshop conclusions and suggest a simple analysis of the performance differences that might be expected between short- and long-pulse sources both of which exploit coupled moderators

  4. 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)

  5. 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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. Development and performance test of small angle neutron spectrometer at HANARO

    International Nuclear Information System (INIS)

    Han, Young Soo; Seong, Baek Seok; Lee, Chang Hee; Lee, Jeong Soo; Hong, Kwang Pyo; Choi, Byung Hoon; Choi, Young Hyun; Shin, Eun Joo; Park, Kook Nam

    2004-12-01

    The construction of Small Angle Neutron Spectrometer(SANS) at the CN beam port in HANARO was completed and has been opened to users in July 2001. the 2-D PSD (two dimensional position sensitive detector), the NVS (neutron velocity selector), the detector chamber rotation system, the detector horizontal moving system, the stepping motors, the beam shutter and the attenuator were fully tested and installed. The performance test of all the components was also completed. Wavelengths and resolutions of the neutron beam monochromatized by the NVS were calibrated using both the time-of-flight method and the diffraction measurement on standard material, the silver behenate. The relationship between the selector speed U[rpm] and the neutron wavelength λ[A] was obtained as λ[A]=0.11077+107171/U[rpm]. The controllers for the sample environments, the beam shutter and the stepping motors were constructed and its control programs for those controllers were also developed. The Beam test for the SANS has been finished and the characteristics of neutron beam was analyzed. The experimental methods of SANS and its data treatment method were established. The performance test of the HANARO SANS compared with that of foreign SANS's. shows that the HANARO SANS is quite well comparable with foreign SANS facilities

  8. 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.

  9. Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor

    International Nuclear Information System (INIS)

    Williams, R.E.; Kopetka, P.; Cook, J.C.; Rowe, J.M.

    2003-01-01

    On March 6, 2002, the NIST Research Reactor resumed routine operation following a six-month shutdown for facility upgrades and maintenance. During the shutdown, the original liquid hydrogen cold neutron source was removed, and the advanced cold source was installed. An optical filter was installed on one of the neutron guides, NG-3, replacing a crystal filter for the 30-m SANS instrument and the guide used between the chopper disks of the Disk Chopper time-of-flight Spectrometer (DCS) installed on NG-4 has been recently reconfigured. Additional improvements in the neutron optics of various instruments are being made. The advanced liquid hydrogen cold neutron source performs as expected, nearly doubling the flux available to most instruments. The measured gains range from about 1.4 at 2 A, to over a factor of two at 15 A. Also as expected, the heat load in the new source increased to 1200 watts, but the previously existing refrigerator has easily accommodated the increase. With intensity gains of a factor of two in the important long wavelength region of the spectrum, the advanced cold source significantly enhances the measurement capability of the cold neutron scattering instrumentation at NIST. The optical filter on NG-3 is also very successful; the 30-m SANS has an additional gain of two at 17 A. A system of refracting lenses and prisms near the SANS sample position has made possible measurements at low Q (0.0005 A -1 ) that were previously not feasible. The DCS has also seen additional intensity gain factors in excess of two for the majority of experiments and at short neutron wavelengths the gains exceed three. In addition, two new triple axis spectrometers will feature double-focusing monochromators in order to exploit the full size of the available thermal and cold neutron beam tubes. The success of the advanced cold source and enhanced neutron optics contributed to the recognition of the NIST Center for Neutron Research as 'the premiere neutron scattering

  10. Evaluation of three partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Rack, H.J.; Pearson, H.S.

    1981-02-01

    The thermal stability and mechanical behavior of three partially volatile candidate neutron shield materials have been evaluated. The results indicate that silicone based rubbers, impregnated with elemental boron or boron carbide, Boro-silicone 236 and Bisco NS-I respectively are more thermally stable than are borated beechwoods, e.g., Permali JN. Mechanical property measurements indicated however that the compressive strength of the borated beechwood is 10 to 48 times higher than that of the silicone-based rubbers. The compressive strengths of the borated beechwood and boron carbide impregnated silicone rubber were substantially more sensitive to test temperature than was the compressive strength of the boron impregnated silicone rubber. Finally the compressive strengths and energy absorbing capability of the boron impregnated silicone rubber is not affected by prior thermal exposure at 425 0 K for 1000h

  11. Multi-target determination of organic ultraviolet absorbents in organism tissues by ultrasonic assisted extraction and ultra-high performance liquid chromatography-tandem mass spectrometry.

    Science.gov (United States)

    Peng, Xianzhi; Jin, Jiabin; Wang, Chunwei; Ou, Weihui; Tang, Caiming

    2015-03-06

    A sensitive and reliable method was developed for multi-target determination of 13 most widely used organic ultraviolet (UV) absorbents (including UV filters and UV stabilizers) in aquatic organism tissues. The organic UV absorbents were extracted using ultrasonic-assisted extraction, purified via gel permeation chromatography coupled with silica gel column chromatography, and determined by ultra-high performance liquid chromatography-tandem mass spectrometry. Recoveries of the UV absorbents from organism tissues mostly ranged from 70% to 120% from fish filet with satisfactory reproducibility. Method quantification limits were 0.003-1.0ngg(-1) dry weight (dw) except for 2-ethylhexyl 4-methoxycinnamate. This method has been applied to analysis of the UV absorbents in wild and farmed aquatic organisms collected from the Pearl River Estuary, South China. 2-Hydroxy-4-methoxybenzophenone and UV-P were frequently detected in both wild and farmed marine organisms at low ngg(-1)dw. 3-(4-Methylbenzylidene)camphor and most of the benzotriazole UV stabilizers were also frequently detected in maricultured fish. Octocrylene and 2-ethylhexyl 4-methoxycinnamate were not detected in any sample. This work lays basis for in-depth study about bioaccumulation and biomagnification of the UV absorbents in marine environment. Copyright © 2015 Elsevier B.V. All rights reserved.

  12. Improved performances of 36 m small-angle neutron scattering spectrometer BATAN in Serpong Indonesia

    International Nuclear Information System (INIS)

    Putra, Edy Giri Rachman; Bharoto; Santoso, Eddy; Ikram, Abarrul

    2009-01-01

    SMARTer, a 36 m small-angle neutron scattering (SANS) spectrometer owned by the National Nuclear Energy Agency of Indonesia (BATAN) was installed at the Neutron Scattering Laboratory (NSL) in Serpong, Indonesia. Lots of works on replacing, upgrading and improving the control system, experimental methods, data collection and reduction in the last two years have been carried out to optimize the performance of SMARTer. Some standard samples such as silver behenate, monodisperse polystyrene nanoparticle, porous silica and block copolymer PS-PEP film were measured for the inter-laboratory comparison.

  13. Deriving Quantitative Crystallographic Information from the Wavelength-Resolved Neutron Transmission Analysis Performed in Imaging Mode

    Directory of Open Access Journals (Sweden)

    Hirotaka Sato

    2017-12-01

    Full Text Available Current status of Bragg-edge/dip neutron transmission analysis/imaging methods is presented. The method can visualize real-space distributions of bulk crystallographic information in a crystalline material over a large area (~10 cm with high spatial resolution (~100 μm. Furthermore, by using suitable spectrum analysis methods for wavelength-dependent neutron transmission data, quantitative visualization of the crystallographic information can be achieved. For example, crystallographic texture imaging, crystallite size imaging and crystalline phase imaging with texture/extinction corrections are carried out by the Rietveld-type (wide wavelength bandwidth profile fitting analysis code, RITS (Rietveld Imaging of Transmission Spectra. By using the single Bragg-edge analysis mode of RITS, evaluations of crystal lattice plane spacing (d-spacing relating to macro-strain and d-spacing distribution’s FWHM (full width at half maximum relating to micro-strain can be achieved. Macro-strain tomography is performed by a new conceptual CT (computed tomography image reconstruction algorithm, the tensor CT method. Crystalline grains and their orientations are visualized by a fast determination method of grain orientation for Bragg-dip neutron transmission spectrum. In this paper, these imaging examples with the spectrum analysis methods and the reliabilities evaluated by optical/electron microscope and X-ray/neutron diffraction, are presented. In addition, the status at compact accelerator driven pulsed neutron sources is also presented.

  14. RADIATION PERFORMANCE OF GAN AND INAS/GAAS QUANTUM DOT BASED DEVICES SUBJECTED TO NEUTRON RADIATION

    Directory of Open Access Journals (Sweden)

    Dhiyauddin Ahmad Fauzi

    2017-05-01

    Full Text Available In addition to their useful optoelectronics functions, gallium nitride (GaN and quantum dots (QDs based structures are also known for their radiation hardness properties. With demands on such semiconductor material structures, it is important to investigate the differences in reliability and radiation hardness properties of these two devices. For this purpose, three sets of GaN light-emitting diode (LED and InAs/GaAs dot-in-a well (DWELL samples were irradiated with thermal neutron of fluence ranging from 3×1013 to 6×1014 neutron/cm2 in PUSPATI TRIGA research reactor. The radiation performances for each device were evaluated based on the current-voltage (I-V and capacitance-voltage (C-V electrical characterisation method. Results suggested that the GaN based sample is less susceptible to electrical changes due to the thermal neutron radiation effects compared to the QD based sample.

  15. Realizing high-performance metamaterial absorber based on the localized surface plasmon resonance in the terahertz regime

    Science.gov (United States)

    Yunfeng, Lin; Xiaoqi, Hu; Lin, Hu

    2018-04-01

    A composite structure design metamaterial absorber is designed and simulated. The proposed composite structure consists of a double-hole sub-structure and a double-metallic particle sub-structure. The damping constant of bulk gold layer is optimized to eliminate the adverse effects of the grain boundary and the surface scattering of thin films on the absorption property. Two absorption peaks (A1 = 58%, A2 = 23%) are achieved based on the localized surface plasmon (LSP) modes resonance. Moreover, the plasmonic hybridization phenomenon between LSP modes is found, which leads to the absorption enhancement between two absorption peaks. The proposed metamaterial absorber holds the property of wide-angle incidence.

  16. Flux Gain for Next-Generation Neutron-Scattering Instruments Resulting From Improved Supermirror Performance

    International Nuclear Information System (INIS)

    Rehm, C.

    2001-01-01

    Next-generation spallation neutron source facilities will offer instruments with unprecedented capabilities through simultaneous enhancement of source power and usage of advanced optical components. The Spallation Neutron Source (SNS), already under construction at Oak Ridge National Laboratory and scheduled to be completed by 2006, will provide greater than an order of magnitude more effective source flux than current state-of-the-art facilities, including the most advanced research reactors. An additional order of magnitude gain is expected through the use of new optical devices and instrumentation concepts. Many instrument designs require supermirror (SM) neutron guides with very high critical angles for total reflection. In this contribution, they discuss how the performance of modern neutron scattering instruments depends on the efficiency of these supermirrors. They outline ideas for enhancing the performance of the SM coatings, particularly for improving the reflectivity at the position of the critical wave vector transfer. A simulation program has been developed which allows different approaches for SM designs to be studied. Possible instrument performance gains are calculated for the example of the SNS reflectometer

  17. The performance of neutron scattering spectrometers at a long-pulse spallation source

    International Nuclear Information System (INIS)

    Pynn, R.

    1997-01-01

    In this document the author considers the performance of a long pulse spallation source for those neutron scattering experiments that are usually performed with a monochromatic beam at a continuous wave (CW) source such as a nuclear reactor. The first conclusion drawn is that comparison of the performance of neutron scattering spectrometers at CW and pulsed sources is simpler for long-pulsed sources than it is for the short-pulse variety. Even though detailed instrument design and assessment will require Monte Carlo simulations (which have already been performed at Los Alamos for SANS and reflectometry), simple arguments are sufficient to assess the approximate performance of spectrometers at an LPSS and to support the contention that a 1 MW long-pulse source can provide attractive performance, especially for instrumentation designed for soft-condensed-matter science. Because coupled moderators can be exploited at such a source, its time average cold flux is equivalent to that of a research reactor with a power of about 15 MW, so only a factor of 4 gain from source pulsing is necessary to obtain performance that is comparable with the ILL. In favorable cases, the gain from pulsing can be even more than this, approaching the limit set by the peak flux, giving about 4 times the performance of the ILL. Because of its low duty factor, an LPSS provides the greatest performance gains for relatively low resolution experiments with cold neutrons. It should thus be considered complementary to short pulse sources which are most effective for high resolution experiments using thermal or epithermal neutrons

  18. The performance of neutron scattering spectrometers at a long-pulse spallation source

    International Nuclear Information System (INIS)

    Pynn, R.

    1995-01-01

    The first conclusion the author wants to draw is that comparison of the performance of neutron scattering spectrometers at CW and pulsed sources is simpler for long-pulsed sources than it is for the short-pulse variety. Even though detailed instrument design and assessment will require Monte Carlo simulations (which have already been performed at Los Alamos for SANS and reflectometry), simple arguments are sufficient to assess the approximate performance of spectrometers at an LPSS and to support the contention that a 1 MW long-pulse source can provide attractive performance, especially for instrumentation designed for soft-condensed-matter science. Because coupled moderators can be exploited at such a source, its time average cold flux is equivalent to that of a research reactor with a power of about 15 MW, so only a factor of 4 gain from source pulsing is necessary to obtain performance that is comparable with the ILL. In favorable cases, the gain from pulsing can be even more than this, approaching the limit set by the peak flux, giving about 4 times the performance of the ILL. Because of its low duty factor, an LPSS provides the greatest performance gains for relatively low resolution experiments with cold neutrons. It should thus be considered complementary to short pulse sources which are most effective for high resolution experiments using thermal or epithermal neutrons

  19. The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

    International Nuclear Information System (INIS)

    Fujita, E.K.; Wade, D.C.

    1990-01-01

    This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

  20. Performance test of micro-fission chambers for in-vessel neutron monitoring of ITER

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Morimoto, Yuichi; Hori, Jun-ichi; Ebisawa, Katsuyuki; Kasai, Satoshi

    2002-03-01

    A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were fabricated and the performance was tested. They are designed to be installed inside the vacuum vessel of the compact ITER (ITER-FEAT) for neutron monitoring. The vacuum leak rate of the dummy chamber with MI cable, resistances of chambers between central conductor and outer sheath, and mechanical strength up to 50G acceleration were confirmed to meet the design criteria. The gamma-ray sensitivity was measured for the dummy chamber with the 60 Co gamma-ray irradiation facility at JAERI Takasaki. The output signals for gamma-rays in Campbelling mode were estimated to be less than 0.1% of those by neutrons at the location behind the blanket module in ITER-FEAT. The detector response for 14 MeV neutrons was investigated with the FNS facility. Excellent linearity between count rates, square of Campbelling voltage and neutron fluxes was confirmed in the temperature range from 20degC (room) to 250degC. However, a positive dependence of 14 MeV neutron count rates on temperature was observed, which might be caused by the increase in the pulse height with temperature rise. Effects of a change of surrounding materials were evaluated by the sensitivity measurements of the micro-fission chamber inserted into the shielding blanket mock-up. The sensitivity was enhanced by slow-downed neutrons, which agreed with the calculation result by MCNP-4C code. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER-FEAT. (author)

  1. Transmission of Thermal Neutrons through Boral

    Energy Technology Data Exchange (ETDEWEB)

    Aakerhielm, F

    1960-06-15

    Transmission measurements have been performed using Maxwellian distributed neutrons from the R1 reactor perpendicularly incident upon a boral absorption plate. American, English, German, Swedish and Swiss samples have been investigated and the results are compared to calculated values. The influence of the absorber grain size is discussed.

  2. Transmission of Thermal Neutrons through Boral

    International Nuclear Information System (INIS)

    Aakerhielm, F.

    1960-06-01

    Transmission measurements have been performed using Maxwellian distributed neutrons from the R1 reactor perpendicularly incident upon a boral absorption plate. American, English, German, Swedish and Swiss samples have been investigated and the results are compared to calculated values. The influence of the absorber grain size is discussed

  3. 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)

  4. Pulsed thermal neutron source at the fast neutron generator.

    Science.gov (United States)

    Tracz, Grzegorz; Drozdowicz, Krzysztof; Gabańska, Barbara; Krynicka, Ewa

    2009-06-01

    A small pulsed thermal neutron source has been designed based on results of the MCNP simulations of the thermalization of 14 MeV neutrons in a cluster-moderator which consists of small moderating cells decoupled by an absorber. Optimum dimensions of the single cell and of the whole cluster have been selected, considering the thermal neutron intensity and the short decay time of the thermal neutron flux. The source has been built and the test experiments have been performed. To ensure the response is not due to the choice of target for the experiments, calculations have been done to demonstrate the response is valid regardless of the thermalization properties of the target.

  5. Studies Performed in Preparation for the Spallation Neutron Source Accumulator Ring Commissioning

    CERN Document Server

    Cousineau, Sarah M; Henderson, Stuart; Holmes, Jeffrey Alan; Plum, Michael

    2005-01-01

    The Spallation Neutron Source accumulator ring will compress 1.5?1014, 1 GeV protons from a 1 ms bunch train to a single 695 ns proton bunch for use in neutron spallation. Due to the high beam power, unprecedented control of beam loss will be required in order to control radiation and allow for hands-on maintenance in most areas of the ring. A number of detailed investigations have been performed to understand the primary sources of beam loss and to predict and mitigate problems associated with radiation hot spots in the ring. The ORBIT particle tracking code is used to perform realistic simulations of the beam accumulation in the ring, including detailed modeling of the injection system, transport through the measured magnet fields including higher order multipoles, and beam loss and collimation. In this paper we present the results of a number of studies performed in preparation for the 2006 commissioning of the accumulator ring.

  6. Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

    2007-07-01

    The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (author)

  7. EJ-309 pulse shape discrimination performance with a high gamma-ray-to-neutron ratio and low threshold

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, A.C., E-mail: Alexis.C.Kaplan@gmail.com [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48104 (United States); Nuclear Engineering and Nonproliferation Division, Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, M.; Enqvist, A.; Dolan, J.L.; Pozzi, S.A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48104 (United States)

    2013-11-21

    Measuring neutrons in the presence of high gamma-ray fluence is a challenge with multi-particle detectors. Organic liquid scintillators such as the EJ-309 are capable of accurate pulse-shape discrimination (PSD) but the chance for particle misclassification is not negligible for some applications. By varying the distance from an EJ-309 scintillator to a strong-gamma-ray source and keeping a weak-neutron source at a fixed position, various gamma-to-neutron ratios can be measured and PSD performance can be quantified. Comparing neutron pulse-height distributions allows for pulse-height specific PSD evaluation, and quantification and visualization of deviation from {sup 252}Cf alone. Even with the addition of the misclassified gamma-rays, the PSD is effective in separating particles so that neutron count rate can be predicted with less than 10% error up to a gamma-to-neutron ratio of almost 650. For applications which can afford a reduction in neutron detection efficiency, PSD can be sufficiently effective in discriminating particles to measure a weak neutron source in a high gamma-ray background. -- Highlights: •We measure neutrons in a high photon background with EJ-309 liquid scintillators. •A low threshold is used to test the limits of particle discrimination. •A weak neutron signal is detectable with a gamma/neutron ratio as high as 770. •Photon pileup most commonly adds to error in classification of neutrons. •Neutron count rates are within 10% of expected rate under high gamma background.

  8. EJ-309 pulse shape discrimination performance with a high gamma-ray-to-neutron ratio and low threshold

    International Nuclear Information System (INIS)

    Kaplan, A.C.; Flaska, M.; Enqvist, A.; Dolan, J.L.; Pozzi, S.A.

    2013-01-01

    Measuring neutrons in the presence of high gamma-ray fluence is a challenge with multi-particle detectors. Organic liquid scintillators such as the EJ-309 are capable of accurate pulse-shape discrimination (PSD) but the chance for particle misclassification is not negligible for some applications. By varying the distance from an EJ-309 scintillator to a strong-gamma-ray source and keeping a weak-neutron source at a fixed position, various gamma-to-neutron ratios can be measured and PSD performance can be quantified. Comparing neutron pulse-height distributions allows for pulse-height specific PSD evaluation, and quantification and visualization of deviation from 252 Cf alone. Even with the addition of the misclassified gamma-rays, the PSD is effective in separating particles so that neutron count rate can be predicted with less than 10% error up to a gamma-to-neutron ratio of almost 650. For applications which can afford a reduction in neutron detection efficiency, PSD can be sufficiently effective in discriminating particles to measure a weak neutron source in a high gamma-ray background. -- Highlights: •We measure neutrons in a high photon background with EJ-309 liquid scintillators. •A low threshold is used to test the limits of particle discrimination. •A weak neutron signal is detectable with a gamma/neutron ratio as high as 770. •Photon pileup most commonly adds to error in classification of neutrons. •Neutron count rates are within 10% of expected rate under high gamma background

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Diagnosing implosion performance at the National Ignition Facility (NIF) by means of neutron spectrometry

    International Nuclear Information System (INIS)

    Frenje, J.A.; Casey, D.T.; Johnson, M. Gatu; Bionta, R.; Bond, E.J.; Caggiano, J.A.; Cerjan, C.; Edwards, J.; Eckart, M.; Fittinghoff, D.N.; Friedrich, S.; Glenzer, S.; Haan, S.; Hatarik, R.; Hatchett, S.; Jones, O.S.; Glebov, V.Yu.; Knauer, J.P.; Grim, G.; Kilkenny, J.D.

    2013-01-01

    The neutron spectrum from a cryogenically layered deuterium–tritium (dt) implosion at the National Ignition Facility (NIF) provides essential information about the implosion performance. From the measured primary-neutron spectrum (13–15 MeV), yield (Y n ) and hot-spot ion temperature (T i ) are determined. From the scattered neutron yield (10–12 MeV) relative to Y n , the down-scatter ratio, and the fuel areal density (ρR) are determined. These implosion parameters have been diagnosed to an unprecedented accuracy with a suite of neutron-time-of-flight spectrometers and a magnetic recoil spectrometer implemented in various locations around the NIF target chamber. This provides good implosion coverage and excellent measurement complementarity required for reliable measurements of Y n , T i and ρR, in addition to ρR asymmetries. The data indicate that the implosion performance, characterized by the experimental ignition threshold factor, has improved almost two orders of magnitude since the first shot taken in September 2010. ρR values greater than 1 g cm −2 are readily achieved. Three-dimensional semi-analytical modelling and numerical simulations of the neutron-spectrometry data, as well as other data for the hot spot and main fuel, indicate that a maximum hot-spot pressure of ∼150 Gbar has been obtained, which is almost a factor of two from the conditions required for ignition according to simulations. Observed Y n are also 3–10 times lower than predicted. The conjecture is that the observed pressure and Y n deficits are partly explained by substantial low-mode ρR asymmetries, which may cause inefficient conversion of shell kinetic energy to hot-spot thermal energy at stagnation. (paper)

  11. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

    1983-01-01

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  12. 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

  13. Design and performance of vacuum capable detector electronics for linear position sensitive neutron detectors

    International Nuclear Information System (INIS)

    Riedel, R.A.; Cooper, R.G.; Funk, L.L.; Clonts, L.G.

    2012-01-01

    We describe the design and performance of electronics for linear position sensitive neutron detectors. The eight tube assembly requires 10 W of power and can be controlled via digital communication links. The electronics can be used without modification in vacuum. Using a transimpedance amplifier and gated integration, we achieve a highly linear system with coefficient of determinations of 0.9999 or better. Typical resolution is one percent of tube length.

  14. Design and performance of vacuum capable detector electronics for linear position sensitive neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, R.A., E-mail: riedelra@ornl.gov [Oak Ridge National Laboratories, Oak Ridge, TN 37830 (United States); Cooper, R.G.; Funk, L.L.; Clonts, L.G. [Oak Ridge National Laboratories, Oak Ridge, TN 37830 (United States)

    2012-02-01

    We describe the design and performance of electronics for linear position sensitive neutron detectors. The eight tube assembly requires 10 W of power and can be controlled via digital communication links. The electronics can be used without modification in vacuum. Using a transimpedance amplifier and gated integration, we achieve a highly linear system with coefficient of determinations of 0.9999 or better. Typical resolution is one percent of tube length.

  15. A neutron amplifier: prospects for reactor-based waste transmutation

    International Nuclear Information System (INIS)

    Blanovsky, A.

    2004-01-01

    A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)

  16. A comparison of back propagation and Generalized Regression Neural Networks performance in neutron spectrometry.

    Science.gov (United States)

    Martínez-Blanco, Ma Del Rosario; Ornelas-Vargas, Gerardo; Solís-Sánchez, Luis Octavio; Castañeda-Miranada, Rodrigo; Vega-Carrillo, Héctor René; Celaya-Padilla, José M; Garza-Veloz, Idalia; Martínez-Fierro, Margarita; Ortiz-Rodríguez, José Manuel

    2016-11-01

    The process of unfolding the neutron energy spectrum has been subject of research for many years. Monte Carlo, iterative methods, the bayesian theory, the principle of maximum entropy are some of the methods used. The drawbacks associated with traditional unfolding procedures have motivated the research of complementary approaches. Back Propagation Neural Networks (BPNN), have been applied with success in neutron spectrometry and dosimetry domains, however, the structure and learning parameters are factors that highly impact in the networks performance. In ANN domain, Generalized Regression Neural Network (GRNN) is one of the simplest neural networks in term of network architecture and learning algorithm. The learning is instantaneous, requiring no time for training. Opposite to BPNN, a GRNN would be formed instantly with just a 1-pass training on the development data. In the network development phase, the only hurdle is to optimize the hyper-parameter, which is known as sigma, governing the smoothness of the network. The aim of this work was to compare the performance of BPNN and GRNN in the solution of the neutron spectrometry problem. From results obtained it can be observed that despite the very similar results, GRNN performs better than BPNN. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. A comparison of back propagation and generalized regression neural networks performance in neutron spectrometry

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Solis S, L. O.; Castaneda M, R.; Vega C, H. R.

    2015-10-01

    The process of unfolding the neutron energy spectrum has been the subject of research for many years. Monte Carlo, iterative methods, the bayesian theory, the principle of maximum entropy are some of the methods used. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Back Propagation Neural Networks (BPNN), have been applied with success in the neutron spectrometry and dosimetry domains, however, the structure and the learning parameters are factors that contribute in a significant way in the networks performance. In artificial neural network domain, Generalized Regression Neural Network (GRNN) is one of the simplest neural networks in term of network architecture and learning algorithm. The learning is instantaneous, which mean require no time for training. Opposite to BPNN, a GRNN would be formed instantly with just a 1-pass training with the development data. In the network development phase, the only hurdle is to tune the hyper parameter, which is known as sigma, governing the smoothness of the network. The aim of this work was to compare the performance of BPNN and GRNN in the solution of the neutron spectrometry problem. From results obtained can be observed that despite the very similar results, GRNN performs better than BPNN. (Author)

  18. Performance assessment of imaging plates for the JHR transfer Neutron Imaging System

    Science.gov (United States)

    Simon, E.; Guimbal, P. AB(; )

    2018-01-01

    The underwater Neutron Imaging System to be installed in the Jules Horowitz Reactor (JHR-NIS) is based on a transfer method using a neutron activated beta-emitter like Dysprosium. The information stored in the converter is to be offline transferred on a specific imaging system, still to be defined. Solutions are currently under investigation for the JHR-NIS in order to anticipate the disappearance of radiographic films commonly used in these applications. We report here the performance assessment of Computed Radiography imagers (Imaging Plates) performed at LLB/Orphée (CEA Saclay). Several imaging plate types are studied, in one hand in the configuration involving an intimate contact with an activated dysprosium foil converter: Fuji BAS-TR, Fuji UR-1 and Carestream Flex XL Blue imaging plates, and in the other hand by using a prototypal imaging plate doped with dysprosium and thus not needing any contact with a separate converter foil. The results for these imaging plates are compared with those obtained with gadolinium doped imaging plate used in direct neutron imaging (Fuji BAS-ND). The detection performances of the different imagers are compared regarding resolution and noise. The many advantages of using imaging plates over radiographic films (high sensitivity, linear response, high dynamic range) could palliate its lower intrinsic resolution.

  19. A comparison of back propagation and Generalized Regression Neural Networks performance in neutron spectrometry

    International Nuclear Information System (INIS)

    Martínez-Blanco, Ma. del Rosario

    2016-01-01

    The process of unfolding the neutron energy spectrum has been subject of research for many years. Monte Carlo, iterative methods, the bayesian theory, the principle of maximum entropy are some of the methods used. The drawbacks associated with traditional unfolding procedures have motivated the research of complementary approaches. Back Propagation Neural Networks (BPNN), have been applied with success in neutron spectrometry and dosimetry domains, however, the structure and learning parameters are factors that highly impact in the networks performance. In ANN domain, Generalized Regression Neural Network (GRNN) is one of the simplest neural networks in term of network architecture and learning algorithm. The learning is instantaneous, requiring no time for training. Opposite to BPNN, a GRNN would be formed instantly with just a 1-pass training on the development data. In the network development phase, the only hurdle is to optimize the hyper-parameter, which is known as sigma, governing the smoothness of the network. The aim of this work was to compare the performance of BPNN and GRNN in the solution of the neutron spectrometry problem. From results obtained it can be observed that despite the very similar results, GRNN performs better than BPNN.

  20. A comparison of back propagation and generalized regression neural networks performance in neutron spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J. M.; Martinez B, M. R.; Solis S, L. O.; Castaneda M, R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Ramon Lopez Velarde 801, Col. Centro, 98000 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    The process of unfolding the neutron energy spectrum has been the subject of research for many years. Monte Carlo, iterative methods, the bayesian theory, the principle of maximum entropy are some of the methods used. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Back Propagation Neural Networks (BPNN), have been applied with success in the neutron spectrometry and dosimetry domains, however, the structure and the learning parameters are factors that contribute in a significant way in the networks performance. In artificial neural network domain, Generalized Regression Neural Network (GRNN) is one of the simplest neural networks in term of network architecture and learning algorithm. The learning is instantaneous, which mean require no time for training. Opposite to BPNN, a GRNN would be formed instantly with just a 1-pass training with the development data. In the network development phase, the only hurdle is to tune the hyper parameter, which is known as sigma, governing the smoothness of the network. The aim of this work was to compare the performance of BPNN and GRNN in the solution of the neutron spectrometry problem. From results obtained can be observed that despite the very similar results, GRNN performs better than BPNN. (Author)

  1. Performance of IPEN/CNEN-SP Neutron Activation Analysis Laboratory for microelement determinations in proficiency testing

    International Nuclear Information System (INIS)

    Armelin, Maria Jose A.; Saiki, Mitiko; Souza, Gilberto B. de; Nogueira, Ana Rita A.

    2009-01-01

    The performance of Neutron Activation Laboratory, IPEN - CNEN/SP, was evaluated for the Ca, Fe, K, Mn, Na and Zn determinations in animal feed samples for ruminants through a proficiency test (PT) program. This PT program is organized by EMBRAPA Cattle Southeast to evaluate laboratories that analyze animal feed samples. Considering the fractions of satisfactory z-scores (%) of evaluated analytes to determine the laboratories performance, the general performance indicator obtained by IPEN - CNEN/SP ranged from 90 to 95% of the satisfactory results during the period of participation in the evaluation, four years. (author)

  2. Neutron diffraction on CeMnAlD{sub x} (0{<=}x{<=}2.5)

    Energy Technology Data Exchange (ETDEWEB)

    Spatz, P.; Gross, K.; Schlapbach, L. [Fribourg Univ. (Switzerland); Fischer, P.; Fauth, F. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-09-01

    CeMnAl was found to absorb considerable amounts of hydrogen. Part of the totally stored hydrogen is absorbed at low pressures (< 10 mbar). Additional hydrogen can be absorbed and desorbed reversible in a wide pressure range (10 mbar to 10 bar) at room temperature. In order to a better understanding of this new metal-hydride system, we performed neutron diffraction on deuterated CeMnAl samples with different D-concentrations. (author) 1 fig., 2 refs.

  3. Evaluation of Neutron Response of Criticality Accident Alarm System Detector to Quasi-Monoenergetic 24 keV Neutrons

    Science.gov (United States)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses.

  4. Evaluation of neutron response of criticality accident alarm system detector to quasi-monoenergetic 24 keV neutrons

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    2016-01-01

    The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses. (author)

  5. Experimental Analysis of the Thermo-Hydraulic Performance on a Cylindrical Parabolic Concentrating Solar Water Heater with Twisted Tape Inserts in an Absorber Tube

    Science.gov (United States)

    Kumar, Birendra; Nayak, Rajen Kumar; Singh, S. N.

    2018-05-01

    A twisted tape inserted in an absorber tube may be an excellent option to enhance the performance of a cylindrical parabolic concentrating solar collector (CPC). The present work is an experimental study of the flow and heat transfer with and without twisted tape inserts in the absorber tube of a CPC. Results are presented for mass flow rates of water, ṁ=0.0198-0.0525 kg/s, twist ratio, y=5-10 and Reynolds number, Re=2577.46-6785.55. In the present study, we found that the outlet water temperature, collector efficiency and Nusselt number (Nu) are higher in the twisted tapes as compared to those without the twisted tape inserts in the absorber tube of the CPC. For fixed mass flow rate of water ṁ, the To and η increased with the decrease in twist ratio, y, and is higher in lower twist ratio, y=5, of the twisted tapes. The whole experiment was performed at the Indian Institute of Technology (ISM) in Dhanbad, India during the months of March-April 2017. Based on the experimental data, the correlations for the Nu and friction factor were also developed.

  6. Effect of Target Configuration on the Neutronic Performance of the Gas-Cooled ADS

    CERN Document Server

    Biss, K; Shetty, N; Nabbi, R

    2013-01-01

    With the utilization of nuclear energy transuranic elements like Pu, Am and Cm are produced causing high, long term radioactivity and radio toxicity, respectively. To reduce the radiological impact on the environment and to the repository Partitioning and Transmutation is considered as an efficient way. In this respect comprehensive research works are performed at different research institutes worldwide. The results show that the transmutation of TRU is achieved with fast neutrons due to the higher fission probability. Based on Accelerator Driven Systems (ADS) those neutrons are used in a particular system, in which mainly liquid metal eutectic (lead bismuth) is used as coolant. The neutronic performance of an ADS system based on gas cooling was studied in this work by using the simulation tool MCNPX. The usage of the Monte-Carlo method in MCNPX allows the simulation of the physical processes in a 3D-model of the core. In dependence of the spallation target material and design several parameters like the mult...

  7. Aperiodic-metamaterial-based absorber

    Directory of Open Access Journals (Sweden)

    Quanlong Yang

    2017-09-01

    Full Text Available The periodic-metamaterial-based perfect absorber has been studied broadly. Conversely, if the unit cell in the metamaterial-based absorber is arranged aperiodically (aperiodic-metamaterial-based absorber, how does it perform? Inspired by this, here we present a systematic study of the aperiodic-metamaterial-based absorber. By investigating the response of metamaterial absorbers based on periodic, Fibonacci, Thue-Morse, and quasicrystal lattices, we found that aperiodic-metamaterial-based absorbers could display similar absorption behaviors as the periodic one in one hand. However, their absorption behaviors show different tendency depending on the thicknesses of the spacer. Further studies on the angle and polarization dependence of the absorption behavior are also presented.

  8. Comparison of calculations with neutron dosimetry measurements performed at the Oak Ridge Poolside Facility

    Energy Technology Data Exchange (ETDEWEB)

    Maerker, R.E.; Williams, M.L.

    1981-01-01

    The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment.

  9. Comparison of calculations with neutron dosimetry measurements performed at the Oak Ridge Poolside Facility

    International Nuclear Information System (INIS)

    Maerker, R.E.; Williams, M.L.

    1981-01-01

    The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment

  10. Experiment of neutron multiplication in lead

    International Nuclear Information System (INIS)

    Jiang Wenmian; Chen Yuan; Liu Rong; Guo Haiping; Shen Jian

    1994-01-01

    The experiments of neutron multiplication in bulk lead have been performed with a total absorption detector (TAD). A hollow polyethylene sphere is used as neutron moderator and absorber of the TAD, which inner and outer diameters are 56 cm and 138 cm respectively. Slow neutron density in TAD is detected with a 6 Li glass scintillator. For Pb thicknesses of 5, 10, 15, 19.6 and 23.1 cm, the neutron multiplications are 1.301, 1.492, 1.599, 1.713 and 1.745 respectively. Overall experimental error is 2.7%. The calculational neutron multiplications with the 1-D ANISN code and ENDF/B-VI file are agreed with experimental ones within experimental error. Moreover, some factors of systematic error of TAD were investigated experimentally, but obvious factors have not been observed yet. (author)

  11. The performance of prototype position-sensitive neutron detectors on SXD at ISIS

    International Nuclear Information System (INIS)

    Wilson, C.C.

    1989-02-01

    The performance of two position-sensitive neutron detector designed for use on the single crystal diffractometer (SXD) at ISIS is assessed. The two detectors examined were the Anger camera 6 Li-glass scintillator PSD and a prototype fibre-optic encoded PSD based on 6 Li-doped ZnS plastic scintillator. The latter detector is found to be both simpler to fabricate and to produce better results on the evidence to date. A summary of some of the expected science from SXD and the performance of the detectors with respect to this is also given. (author)

  12. 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

  13. Development of time projection chamber for precise neutron lifetime measurement using pulsed cold neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Arimoto, Y. [High Energy Accelerator Research Organization, Ibaraki (Japan); Higashi, N. [Graduate School of Science, University of Tokyo, Tokyo (Japan); Igarashi, Y. [High Energy Accelerator Research Organization, Ibaraki (Japan); Iwashita, Y. [Institute for Chemical Research, Kyoto University, Kyoto (Japan); Ino, T. [High Energy Accelerator Research Organization, Ibaraki (Japan); Katayama, R. [Graduate School of Science, University of Tokyo, Tokyo (Japan); Kitaguchi, M. [Kobayashi-Maskawa Institute, Nagoya University, Aichi (Japan); Kitahara, R. [Graduate School of Science, Kyoto University, Kyoto (Japan); Matsumura, H.; Mishima, K. [High Energy Accelerator Research Organization, Ibaraki (Japan); Nagakura, N.; Oide, H. [Graduate School of Science, University of Tokyo, Tokyo (Japan); Otono, H., E-mail: otono@phys.kyushu-u.ac.jp [Research Centre for Advanced Particle Physics, Kyushu University, Fukuoka (Japan); Sakakibara, R. [Department of Physics, Nagoya University, Aichi (Japan); Shima, T. [Research Center for Nuclear Physics, Osaka University, Osaka (Japan); Shimizu, H.M.; Sugino, T. [Department of Physics, Nagoya University, Aichi (Japan); Sumi, N. [Faculty of Sciences, Kyushu University, Fukuoka (Japan); Sumino, H. [Department of Basic Science, University of Tokyo, Tokyo (Japan); Taketani, K. [High Energy Accelerator Research Organization, Ibaraki (Japan); and others

    2015-11-01

    A new time projection chamber (TPC) was developed for neutron lifetime measurement using a pulsed cold neutron spallation source at the Japan Proton Accelerator Research Complex (J-PARC). Managing considerable background events from natural sources and the beam radioactivity is a challenging aspect of this measurement. To overcome this problem, the developed TPC has unprecedented features such as the use of polyether-ether-ketone plates in the support structure and internal surfaces covered with {sup 6}Li-enriched tiles to absorb outlier neutrons. In this paper, the design and performance of the new TPC are reported in detail.

  14. The new vertical neutron beam line at the CERN n-TOF facility design and outlook on the performance

    Energy Technology Data Exchange (ETDEWEB)

    Weiß, C., E-mail: christina.weiss@cern.ch [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Chiaveri, E.; Girod, S.; Vlachoudis, V.; Aberle, O. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Barros, S. [Instituto Tecnológico e Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Lisboa (Portugal); Bergström, I. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Berthoumieux, E. [Commissariat à l’Énergie Atomique (CEA) Saclay – Irfu, Gif-sur-Yvette (France); Calviani, M. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Guerrero, C.; Sabaté-Gilarte, M. [Universidad de Sevilla (Spain); European Organization for Nuclear Research (CERN), Geneva (Switzerland); Tsinganis, A. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); National Technical University of Athens (NTUA) (Greece); Andrzejewski, J. [Uniwersytet Łódzki, Lodz (Poland); Audouin, L. [Centre National de la Recherche Scientifique/IN2P3 – IPN, Orsay (France); Bacak, M. [Atominstitut, Technische Universität Wien (Austria); Balibrea-Correa, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Barbagallo, M. [Istituto Nazionale di Fisica Nucleare, Bari (Italy); Bécares, V. [Centro de Investigaciones Energeticas Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); and others

    2015-11-01

    At the neutron time-of-flight facility n-TOF at CERN a new vertical beam line was constructed in 2014, in order to extend the experimental possibilities at this facility to an even wider range of challenging cross-section measurements of interest in astrophysics, nuclear technology and medical physics. The design of the beam line and the experimental hall was based on FLUKA Monte Carlo simulations, aiming at maximizing the neutron flux, reducing the beam halo and minimizing the background from neutrons interacting with the collimator or back-scattered in the beam dump. The present paper gives an overview on the design of the beam line and the relevant elements and provides an outlook on the expected performance regarding the neutron beam intensity, shape and energy resolution, as well as the neutron and photon backgrounds.

  15. The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

    Science.gov (United States)

    Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

    2014-08-01

    The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  16. Performance Comparison of Mode-Locked Erbium-Doped Fiber Laser with Nonlinear Polarization Rotation and Saturable Absorber Approaches

    International Nuclear Information System (INIS)

    Ismail, M. A.; Tan, S. J.; Shahabuddin, N. S.; Harun, S. W.; Arof, H.; Ahmad, H.

    2012-01-01

    A mode-locked erbium-doped fiber laser (EDFL) is demonstrated using a highly concentrated erbium-doped fiber (EDF) as the gain medium in a ring configuration with and without a saturable absorber (SA). Without the SA, the proposed laser generates soliton pulses with a repetition rate of 12 MHz, pulse width of 1.11 ps and energy pulse of 1.6 pJ. By incorporating SA in the ring cavity, the optical output of the laser changes from soliton to stretched pulses due to the slight change in the group velocity dispersion. With the SA, a cleaner pulse is obtained with a repetition rate of 11.3 MHz, a pulse width of 0.58 ps and a pulse energy of 2.3 pJ. (fundamental areas of phenomenology(including applications))

  17. Improvement of the performance of microwave X band absorbers based on pure and doped Ba-hexaferrite

    Energy Technology Data Exchange (ETDEWEB)

    Seyyed Afghahi, Seyyed Salman [Department of Materials Science and Engineering, Imam Hossein University, Tehran (Iran, Islamic Republic of); Jafarian, Mojtaba, E-mail: m.jafarian@srbiau.ac.ir [Young Researchers and Elite Club, Science and Research Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Salehi, Mohsen [Young Researchers and Elite Club, Science and Research Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Atassi, Yomen [Department of Applied Physics, Higher Institute for Applied Sciences and Technology, Damascus (Syrian Arab Republic)

    2017-01-01

    Optimum Fe/Ba mole ratio is determined in doped Ba-hexaferrite prepared via mechanical activation. X-ray diffractometer (XRD), scanning electron microscope (SEM), vibrating sample magnetometer (VSM) and vector network analyzer are used to analyze phases, structures, electromagnetic and microwave absorption properties. The mole ratio of Fe/Ba=10 is detected to be optimum for doping and synthesizing the Ba-hexaferrite. In order to achieve high absorption in X band the ions of Zr{sup 4+}–Sn{sup 4+}–Ti{sup 4+}–M{sup 2+} (M=Mg{sup 2+}, Zn{sup 2+}, Cu{sup 2+}, Co{sup 2+}) are used as dopants. The results indicate the formation of single phase Ba-hexaferrite in either pure or doped compounds without any non-magnetic intermediate phases and with spherical and hexagonal morphologies respectively for the pure and doped ferrite. It is found out that BaCo{sub 2}Zr(SnTi){sub 0.5}Fe{sub 8}O{sub 19} compound has the maximum saturation magnetization (49.80 emu/g). Also the composite of BaCo{sub 2}Zr(SnTi){sub 0.5}Fe{sub 8}O{sub 19} 50 wt% in epoxy resin exhibits a minimum reflection loss of −29 dB at 12.2 GHz with 2.6 GHz bandwidth. - Highlights: • Optimum Fe/Ba mole ratio is determined in doped Ba-hexaferrite. • Formation of a single phase M-hexaferrite, pure or doped when Fe/Ba=10. • Preparation of microwave absorber of BaCo{sub 2}Zr(SnTi){sub 0.5}Fe{sub 8}O{sub 19}/epoxy resin, 50 wt%. • The absorber exhibits a minimum RL of −29 dB at 12.2 GHz with 2.6 GHz bandwidth.

  18. Octalithium plumbate as breeding blanket ceramic: Neutronic performances, synthesis and partial characterization

    International Nuclear Information System (INIS)

    Colominas, S.; Palermo, I.; Abellà, J.; Gómez-Ros, J.M.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Definition of a suitable configuration for the Li 8 PbO 6 breeding blanket design. ► Demonstration of the feasibility of Li 8 PbO 6 as a breeding material. ► Synthesis optimization in the Li 8 PbO 6 production. ► Characterization of Li 8 PbO 6 by X-ray phase analysis is discussed. - Abstract: A neutronic assessment of the performances of a helium-cooled Li 8 PbO 6 breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li 8 PbO 6 as breeding material. Furthermore, the synthesis and characterization of Li 8 PbO 6 by X-ray phase analysis are also discussed.

  19. Speciation analysis of cobalt in foods by high-performance liquid chromatography and neutron activation analysis

    International Nuclear Information System (INIS)

    Muto, Toshio; Koyama, Motoko

    1994-01-01

    A combined method by coupling high-performance liquid chromatography (HPLC, as a separation method) with neutron activation analysis (as a detection method) have been applied to the speciation analysis of cobalt in daily foods (e.g. egg, fish and milk). Cobalt species including free cobalt, vitamin B 12 and protein-bound cobalt were separated with a preparative HPLC and a centrifuge. Subsequently, the determination of cobalt in the separated species was made by neutron activation analysis. The results showed that the content of the total cobalt in the foods was found to lie in the range 0.4-11ng/g(0.4-11ppb) based on wet weight. The compositions of free cobalt, vitamin B 12 and protein-bound cobalt were ranged 16-43%, 55-73%, 2.3-17%, respectively. These experimental evidences suggest that the combination of HPLC and neutron activation analysis is expected to be a useful tool for speciation analysis of trace elements in biological as well as environmental materials. (author)

  20. FPGA hardware acceleration for high performance neutron transport computation based on agent methodology - 318

    International Nuclear Information System (INIS)

    Shanjie, Xiao; Tatjana, Jevremovic

    2010-01-01

    The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in developed countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor geometries. For the first time this approach is applied to accelerate the neutronics analysis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with data flow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency. (authors)

  1. Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, Sean [Transatomic Power Corp., Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corp., Cambridge, MA (United States); Massie, Mark [Transatomic Power Corp., Cambridge, MA (United States); Davidson, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

  2. Performance of the upgraded ultracold neutron source at Los Alamos National Laboratory and its implication for a possible neutron electric dipole moment experiment

    Science.gov (United States)

    Ito, T. M.; Adamek, E. R.; Callahan, N. B.; Choi, J. H.; Clayton, S. M.; Cude-Woods, C.; Currie, S.; Ding, X.; Fellers, D. E.; Geltenbort, P.; Lamoreaux, S. K.; Liu, C.-Y.; MacDonald, S.; Makela, M.; Morris, C. L.; Pattie, R. W.; Ramsey, J. C.; Salvat, D. J.; Saunders, A.; Sharapov, E. I.; Sjue, S.; Sprow, A. P.; Tang, Z.; Weaver, H. L.; Wei, W.; Young, A. R.

    2018-01-01

    The ultracold neutron (UCN) source at Los Alamos National Laboratory (LANL), which uses solid deuterium as the UCN converter and is driven by accelerator spallation neutrons, has been successfully operated for over 10 years, providing UCN to various experiments, as the first production UCN source based on the superthermal process. It has recently undergone a major upgrade. This paper describes the design and performance of the upgraded LANL UCN source. Measurements of the cold neutron spectrum and UCN density are presented and compared to Monte Carlo predictions. The source is shown to perform as modeled. The UCN density measured at the exit of the biological shield was 184 (32 ) UCN /cm3 , a fourfold increase from the highest previously reported. The polarized UCN density stored in an external chamber was measured to be 39 (7 ) UCN /cm3 , which is sufficient to perform an experiment to search for the nonzero neutron electric dipole moment with a one-standard-deviation sensitivity of σ (dn) =3 ×10-27e cm .

  3. Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

    International Nuclear Information System (INIS)

    Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

    2002-01-01

    An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

  4. Comparison of quality assurance for performance and safety characteristics of the facility for Boron Neutron Capture therapy in Petten/NL with medical electron accelerators

    International Nuclear Information System (INIS)

    Rassow, Juergen; Stecher-Rasmussen, Finn; Voorbraak, Wim; Moss, Ray; Vroegindeweij, Corine; Hideghety, Katalin; Sauerwein, Wolfgang

    2001-01-01

    Background and purpose: The European Council Directive on health protection 97/43/EURATOM requires radiotherapy quality assurance programmes for performance and safety characteristics including acceptance and repeated tests. For Boron Neutron Capture therapy (BNCT) at the High Flux Reactor (HFR) in Petten/NL such a programme has been developed on the basis of IEC publications for medical electron accelerators. Results: The fundamental differences of clinical dosimetry for medical electron accelerators and BNCT are presented and the order of magnitude of dose components and their stability and that of the main other influencing parameter 10 B concentration for BNCT patient treatments. A comparison is given for requirements for accelerators and BNCT units indicating items which are not transferable, equal or additional. Preliminary results of in vivo measurements done with a set of 55 Mn, 63 Cu and 197 Au activation foils for all single fields for the four fractions at all 15 treated patients show with <±4% up to now a worse reproducibility than the used dose monitoring systems (±1.5%) caused by influence of hair position on the foil-skull distance. Conclusions: Despite the more complex clinical dosimetry (because of four relevant dose components, partly of different linear energy transfer (LET)) BNCT can be regulated following the principles of quality assurance procedures for therapy with medical electron accelerators. The reproducibility of applied neutron fluence (proportional to absorbed doses) and the main safety aspects are equal for all teletherapy methods including BNCT

  5. The Design and Performance of the Spallation Neutron Source Low-Level RF Control System

    CERN Document Server

    Champion, M; Kasemir, K; Ma, H; Piller, C

    2004-01-01

    The Spallation Neutron Source linear accelerator low-level RF control system has been developed within a collaboration of Lawrence Berkeley, Los Alamos, and Oak Ridge national laboratories. Three distinct generations of the system, described in a previous publication [1], have been used to support beam commissioning at Oak Ridge. The third generation system went into production in early 2004, with installation in the coupled-cavity and superconducting linacs to span the remainder of the year. The final design of this system will be presented along with results of performance measurements.

  6. Experimental demonstration of H∞ filter performance for dynamic compensation of rhodium neutron detectors

    International Nuclear Information System (INIS)

    Park, Moon-Ghu; Choi, Yu-Sun; Lee, Kwang-Dae

    2008-01-01

    This paper describes the experimental demonstration of the theoretical result of the previous work on LMI (linear matrix inequality) based H ∞ filter for time-delay compensation of self-powered neutron detectors. The filter gains are optimized in the sense of noise attenuation level of H ∞ setting. By introducing bounded real lemma, the conventional algebraic Riccati inequalities are converted into linear matrix inequalities (LMIs). Finally, the filter design problem is solved via the convex optimization framework using LMIs. The experimental measurements of rhodium detector signal from a research reactor show that the predicted theoretical filter performance is verified by showing successful reconstruction of the reference power signal

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

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  8. Performance of the second Deep Inelastic Neutron Scatering spectrometer at the Bariloche electron LINAC

    International Nuclear Information System (INIS)

    Palomino, L A Rodríguez; Blostein, J J; Dawidowski, J

    2013-01-01

    We report on the new Deep Inelastic Neutron Scattering detector bank recently implemented at the Bariloche electron LINAC. We show the characterization and calibration process carried out, which comprises the determinarion of the detector bank efficiency, and the evaluation of the performance of the filter difference technique. As part of the benchmarking process, polyethylene spectra were measured and analyzed, and the scattering cross sections for carbon and hydrogen were determined in the process. With the addition of this new detector bank to the existing one, we evaluate the combined capacity of the two banks

  9. Absorbance detector for high performance liquid chromatography based on a deep-UV light-emitting diode at 235nm.

    Science.gov (United States)

    da Silveira Petruci, João Flavio; Liebetanz, Michael G; Cardoso, Arnaldo Alves; Hauser, Peter C

    2017-08-25

    In this communication, we describe a flow-through optical absorption detector for HPLC using for the first time a deep-UV light-emitting diode with an emission band at 235nm as light source. The detector is also comprised of a UV-sensitive photodiode positioned to enable measurement of radiation through a flow-through cuvette with round aperture of 1mm diameter and optical path length of 10mm, and a second one positioned as reference photodiode; a beam splitter and a power supply. The absorbance was measured and related to the analyte concentration by emulating the Lambert-Beer law with a log-ratio amplifier circuitry. This detector showed noise levels of 0.30mAU, which is comparable with our previous LED-based detectors employing LEDs at 280 and 255nm. The detector was coupled to a HPLC system and successfully evaluated for the determination of the anti-diabetic drugs pioglitazone and glimepiride in an isocratic separation and the benzodiazepines flurazepam, oxazepam and clobazam in a gradient elution. Good linearities (r>0.99), a precision better than 0.85% and limits of detection at sub-ppm levels were achieved. Copyright © 2017 Elsevier B.V. All rights reserved.

  10. Absorbable and biodegradable polymers

    CERN Document Server

    Shalaby, Shalaby W

    2003-01-01

    INTRODUCTION NOTES: Absorbable/Biodegradable Polymers: Technology Evolution. DEVELOPMENT AND APPLICATIONOF NEW SYSTEMS: Segmented Copolyesters with Prolonged Strength Retention Profiles. Polyaxial Crystalline Fiber-Forming Copolyester. Polyethylene Glycol-Based Copolyesters. Cyanoacrylate-Based Systems as Tissue Adhesives. Chitosan-Based Systems. Hyaluronic Acid-Based Systems. DEVELOPMENTS IN PREPARATIVE, PROCESSING, AND EVALUATION METHODS: New Approaches to the Synthesis of Crystalline. Fiber-Forming Aliphatic Copolyesters. Advances in Morphological Development to Tailor the Performance of Me

  11. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  12. Performance Evaluation of the Neutron Coincidence Counter for the Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

    Lee, S.Y.; Li, T.K.; Menlove, Howard O.; Kim, H.D.; Ko, W.I.; Park, S.W.

    2005-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

  13. Performance test of a gamma/neutron mapper on stored TRU waste durms at the RWMC

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Josten, N.E.; Lawrence, R.S.

    1995-01-01

    The results from a performance test of a γ- and neutron-radiation measurement instrument used to provide two-dimensional radiation field maps are reported. The performance test was conducted at the Transuranic Storage Area of the Radioactive Waste Management Complex (RWMC) where interim storage is provided for 55-gal. drums of TRU waste from the Department of Energy's Rocky Flats Plant. The performance test consisted of scanning drums stacked five high and five wide to identify high radiation areas and possible discrepancies with the waste manifest. Scans were taken at standoff distances of 15 cm, 30 cm, 45 cm and 90 cm. Data were acquired at scan speeds of 7.5 cm/s and 15 cm/s. The results of these scans are presented as one, two and three dimensional contour plots of the radiation fields. A comparison of these results with manifests of these drums are compared and discussed. While the T-radiation fields as measured by the Health Physicist and by the radiation maps are in general in agreement, the TRU content as given in the manifest did not often correlate with the neutron map

  14. Advances in absorbed dose measurement standards at the australian radiation laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boas, J.F.; Hargrave, N.J.; Huntley, R.B.; Kotler, L.H.; Webb, D.V.; Wise, K.N. [Australian Radiation Laboratory, Yallambie, VIC (Australia)

    1996-12-31

    The applications of ionising radiation in the medical and industrial fields require both an accurate knowledge of the amount of ionising radiation absorbed by the medium in question and the capability of relating this to National and International standards. The most useful measure of the amount of radiation is the absorbed dose which is defined as the energy absorbed per unit mass. For radiotherapy, the reference medium is water, even though the measurement of the absorbed dose to water is not straightforward. Two methods are commonly used to provide calibrations in absorbed dose to water. The first is the calibration of the chamber in terms of exposure in a Cobalt-60 beam, followed by the conversion by a protocol into dose to water in this and higher energy beams. The other route is via the use of a graphite calorimeter as a primary standard device, where the conversion from absorbed dose to graphite to absorbed dose in water is performed either by theoretical means making use of cavity ionisation theory, or by experiment where the graphite calorimeter and secondary standard ionisation chamber are placed at scaled distances from the source of the radiation beam (known as the Dose-Ratio method). Extensive measurements have been made at Cobalt-60 at ARL using both the exposure and absorbed dose to graphite routes. Agreement between the ARL measurements and those based on standards maintained by ANSTO and NPL is within {+-} 0.3%. Absorbed dose measurements have also been performed at ARL with photon beams of nominal energy 16 and 19 MeV obtained from the ARL linac. The validity of the protocols at high photon energies, the validity of the methods used to convert from absorbed dose in graphite to absorbed dose in water and the validity of the indices used to specify the beams are discussed. Brief mention will also be made of the establishment of a calibration facility for neutron monitors at ARL and of progress in the development of ERP dosimetry.

  15. Performance of a tagged neutron inspection system (TNIS) based on portable sealed generators

    International Nuclear Information System (INIS)

    Nebbia, G.; Pesente, S.; Lunardon, M.; Viesti, G.; LeTourneur, P.; Heuveline, F.; Mangeard, M.; Tcheng, C.

    2004-01-01

    A portable sealed neutron generator has been modified to produce 14MeV tagged neutron beams with an embedded YAP:Ce scintillation detector. The system has been tested by detecting the coincident gamma-rays produced in the irradiation of a graphite sample by means of a standard NaI(Tl) scintillator. Time resolution of about δt=4-5ns (FWHM) has been measured. The sealed neutron tube has been operated up to 10 7 neutron/s. Possible applications in non-destructive assays and future developments of the Tagged Neutron Inspection System concept are discussed

  16. Mechanical performance optimization of neutron shielding material based on short carbon fiber reinforced B4C/epoxy resin

    International Nuclear Information System (INIS)

    Wang Peng; Tang Xiaobin; Chen Feida; Chen Da

    2013-01-01

    To satisfy engineering requirements for mechanics performance of neutron shielding material, short carbon fiber was used to reinforce the traditional containing B 4 C neutron shielding material and effects of fiber content, length and surface treatment to mechanics performance of material was discussed. Based on Americium-Beryllium neutron source, material's neutron shielding performance was tested. The result of experiment prove that tensile strength of material which the quality ratio of resin and fiber is 5:1 is comparatively excellent for 10wt% B 4 C of carbon fiber reinforced epoxy resin. The tensile properties of material change little with the fiber length ranged from 3-10 mm The treatment of fiber surface with silane coupling agent KH-550 can increase the tensile properties of materials by 20% compared with the untreated of that. A result of shielding experiment that the novel neutron shielding material can satisfy the neutron shielding requirements can be obtained by comparing with B 4 C/polypropylene materials. The material has good mechanical properties and wide application prospect. (authors)

  17. A Simple Correlation for Neutron Capture Rates from Nuclear Masses

    Energy Technology Data Exchange (ETDEWEB)

    Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-01-30

    Recent studies of neutron capture performed at LANL have revealed a previously unrecognized connection between nuclear masses and the average neutron capture cross section. A team of three scientists from Los Alamos (P-27), Yale Univ., and Istanbul Univ. (Turkey) recently discovered this connection and have published their results as a Rapid Communication in Physical Review C. Neutron capture is a reaction in which a free neutron is absorbed by the nucleus, keeping the element unchanged, but changing isotopes. This reaction is typically exothermic. As a result, the reaction can proceed even when many other reaction channels are closed. In an astrophysical environment, this means that neutron capture is the primary mechanism by which all of the elements with atomic number greater than nickel are produced is neutron capture.

  18. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  19. Baseline drift effect on the performance of neutron and γ ray discrimination using frequency gradient analysis

    International Nuclear Information System (INIS)

    Liu Guofu; Luo Xiaoliang; Yang Jun; Lin Cunbao; Hu Qingqing; Peng Jinxian

    2013-01-01

    Frequency gradient analysis (FGA) effectively discriminates neutrons and γ rays by examining the frequency-domain features of the photomultiplier tube anode signal. This approach is insensitive to noise but is inevitably affected by the baseline drift similar to other pulse shape discrimination methods. The baseline drift effect is attributed to factors such as power line fluctuation, dark current, noise disturbances, hum, and pulse tail in front-end electronics. This effect needs to be elucidated and quantified before the baseline shift can be estimated and removed from the captured signal. Therefore, the effect of baseline shift on the discrimination performance of neutrons and γ rays with organic scintillation detectors using FGA is investigated in this paper. The relationship between the baseline shift and discrimination parameters of FGA is derived and verified by an experimental system consisting of an americium—beryllium source, a BC501A liquid scintillator detector, and a 5 GSample/s 8-bit oscilloscope. The theoretical and experimental results both show that the estimation of the baseline shift is necessary, and the removal of baseline drift from the pulse shapes can improve the discrimination performance of FGA. (authors)

  20. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    International Nuclear Information System (INIS)

    Seppaelae, Malla

    2008-01-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  1. Neutronic density perturbation by probes

    International Nuclear Information System (INIS)

    Vigon, M. A.; Diez, L.

    1956-01-01

    The introduction of absorbent materials of neutrons in diffuser media, produces local disturbances of neutronic density. The disturbance depends especially on the nature and size of the absorbent. Approximated equations which relates te disturbance and the distance to the absorbent in the case of thin disks have been drawn. The experimental comprobation has been carried out in two especial cases. In both cases the experimental results are in agreement with the calculated values from these equations. (Author)

  2. The performance of an amorphous silicon flat panel for neutron imaging at the PSI NEUTRA facility

    International Nuclear Information System (INIS)

    Estermann, Mirko; Frei, Gabriel; Lehmann, Eberhard; Vontobel, Peter

    2005-01-01

    Commonly applied imaging methods in neutron radiography use a CCD camera in conjunction with a scintillator or imaging plates. CCDs and imaging plates have desirable properties concerning resolution, linearity, dynamic range and signal-to-noise ratio (S/N) but both techniques have the disadvantage of a slow read out and for the CCD camera, an additional disadvantage is the loss of light through the optical system. Amorphous silicon detectors, originally developed for medical and industrial X-ray imaging, generally do not have the above-mentioned disadvantages. These detectors have a much faster readout and, in comparison to the generally used crystalline silicon, can be put directly in the X-ray or neutron beam without being damaged. This type of detector also does not require any optical interface, minimizing possible light loss. The detector is operated at room temperature, which has some influence on the noise. Using the whole dynamic range with a low gain, results in a S/N of up to 30, for normal applications, however, a S/N of about 15-20 is typical. The main drawback of this imaging device is the dynamic range of only 12 bits and the relatively complicated operating system in which different operation modes can be chosen. In 2003, successful experiments were performed with this new device, but it is still in its fledgling stages and improvements from the manufacturer as well as the experience from the NEUTRA team will help to advance this technique for neutron imaging in a most efficient way

  3. The TOFOR Neutron Spectrometer For High-Performance Measurements of D Plasma Fuel Ion Properties

    International Nuclear Information System (INIS)

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

    2008-01-01

    The impact of scattered neutrons on the total flux reaching the TOFOR spectrometer at JET has been studied to allow for improvement of the data analysis. The scattered neutrons are demonstrated to contribute significantly to the flux. This will have implications for any neutron diagnostic on ITER

  4. Thermal Performance of Solar Air Heater Having Absorber Plate with V-Down Discrete Rib Roughness for Space-Heating Applications

    Directory of Open Access Journals (Sweden)

    Rajendra Karwa

    2013-01-01

    Full Text Available The paper presents results of thermal performance analysis of a solar air heater with v-down discrete rib roughness on the air flow side of the absorber plate, which supplies heated air for space heating applications. The air heater operates in a closed loop mode with inlet air at a fixed temperature of 295 K from the conditional space. The ambient temperature varied from 278 K to 288 K corresponding to the winter season of Western Rajasthan, India. The results of the analysis are presented in the form of performance plots, which can be utilized by a designer for calculating desired air flow rate at different ambient temperature and solar insolation values.

  5. Neutronics of pulsed spallation neutron sources

    CERN Document Server

    Watanabe, N

    2003-01-01

    Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

  6. Directional epithermal neutron detector

    International Nuclear Information System (INIS)

    Givens, W.W.; Mills, W.R. Jr.

    1986-01-01

    A borehole tool for epithermal neutron die-away logging of subterranean formations surrounding a borehole is described which consists of: (a) a pulsed source of fast neutrons for irradiating the formations surrounding a borehole, (b) at least one neutron counter for counting epithermal neutrons returning to the borehole from the irradiated formations, (c) a neutron moderating material, (d) an outer thermal neutron shield providing a housing for the counter and the moderating material, (e) an inner thermal neutron shield dividing the housing so as to provide a first compartment bounded by the inner thermal neutron shield and a first portion of the outer thermal neutron shield and a second compartment bounded by the inner thermal neutron shield and a second portion of the outer thermal neutron shield, the counter being positioned within the first compartment and the moderating material being positioned within the second compartment, and (f) means for positioning the borehole tool against one side of the borehole wall and azimuthally orienting the borehole tool such that the first chamber is in juxtaposition with the borehole wall, the formation epithermal neutrons penetrating into the first chamber through the first portion of the outer thermal neutron shield are detected by the neutron counter for die-away measurement, thereby maximizing the directional sensitivty of the neutron counter to formation epithermal neutrons, the borehole fluid epithermal neutrons penetrating into the second chamber through the second chamber through the second portion of the outer thermal neutron shield are largely slowed down and lowered in energy by the moderating material and absorbed by the inner thermal neutron shield before penetrating into the first chamber, thereby minimizing the directional sensitivity of the neutron counter to borehole fluid epithermal neutrons

  7. A novel Fe–Cr–Nb matrix composite containing the TiB{sub 2} neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

    Energy Technology Data Exchange (ETDEWEB)

    Litwa, Przemysław [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Perkowski, Krzysztof [Department of Nanotechnology, Institute of Ceramics and Building Materials, Postępu 9, 02-676 Warsaw (Poland); Zasada, Dariusz [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Kobus, Izabela; Konopka, Gustaw [Department of Nanotechnology, Institute of Ceramics and Building Materials, Postępu 9, 02-676 Warsaw (Poland); Czujko, Tomasz [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Varin, Robert A., E-mail: robert.varin@uwaterloo.ca [Department of Mechanical and Mechatronics Engineering, University of Waterloo, 200 University Ave., Waterloo, ON N2L 3G1 (Canada)

    2016-07-25

    The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB{sub 2} and a small quantity of Y{sub 2}O{sub 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{sub 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{sub 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{sub 2} neutron

  8. A honeycomb-like three-dimensional metamaterial absorber via super-wideband and wide-angle performances at millimeter wave and low THz frequencies

    Science.gov (United States)

    Vahidi, Alireza; Rajabalipanah, Hamid; Abdolali, Ali; Cheldavi, Ahmad

    2018-04-01

    Achieving wideband absorption via three-dimensional (3D) metamaterials has revealed as a new emerging innovative field of research, especially in recent years. Here, a novel 3D metamaterial absorber (MA) having a sixfold symmetry is designed which consists of periodic resistive honeycomb-like units. The proposed 3D MA exhibits a strong absorptivity above 90% in the widest bandwidth ever reported to the authors' knowledge from 50 to 460 GHz (the bandwidth ratio larger than 1:9), covering both millimeter wave and low -terahertz spectra. To understand the physical mechanism of absorption, the electric field and surface current distributions, the power loss density as well as the deteriorating effects of the high-order Floquet modes are monitored and discussed. As a distinctive feature in comparison to the similar 3D MAs, our engineered absorber provides multiple resonances, contributing to further broadening of the operating bandwidth. In addition, it is shown that the honeycomb-like MA retains its polarization-insensitive absorption in a wide range of incident wave angles and polarization angles. Due to flexibility of the design, these superior performances can be simply extended to terahertz, infrared and visible frequencies, potentially leading to many promising applications in imaging, sensing, and camouflage technology.

  9. A contribution to the development of the modular neutron detector (DEMON): performance evaluation through measurements and simulations; Contribution a la realisation du detecteur modulaire de neutrons (DEMON): etudes des performances par mesures et simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mouatassim, S

    1994-07-01

    The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON`s modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite {sup 126}Ba system formed in the {sup 19}F + {sup 107}Ag entrance channel. (from author) 70 figs., 99 refs.

  10. Liquid metal reactor absorber technology

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1990-10-01

    The selection of boron carbide as the reference liquid metal reactor absorber material is supported by results presented for irradiation performance, reactivity worth compatibility, and benign failure consequences. Scram response requirements are met easily with current control rod configurations. The trend in absorber design development is toward larger sized pins with fewer pins per bundle, providing economic savings and improved hydraulic characteristics. Very long-life absorber designs appear to be attainable with the application of vented pin and sodium-bonded concepts. 3 refs., 3 figs

  11. Performance of a reflectometer at continuous wave and pulsed neutron sources

    International Nuclear Information System (INIS)

    Fitzsimmons, M.R.

    1995-01-01

    The Monte-Carlo simulations presented here involve simulations of reflectivity measurements of one sample using a reflectometer of traditional geometry at different neutron sources. The same reflectometer was used in all simulations. Only the characteristics of the neutron source, and the technique used to measure neutron wavelength were changed. In the case of the CW simulation, a monochromating crystal was used to select a nearly monochromatic beam (MB) from the neutron spectrum. In the simulations of the pulse sources, the time needed to traverse a fixed distance was measured, from which neutron wavelength is deduced

  12. Performance analysis of absorption heat transformer cycles using ionic liquids based on imidazolium cation as absorbents with 2,2,2-trifluoroethanol as refrigerant

    International Nuclear Information System (INIS)

    Ayou, Dereje S.; Currás, Moisés R.; Salavera, Daniel; García, Josefa; Bruno, Joan C.; Coronas, Alberto

    2014-01-01

    Highlights: • TFE + [emim][BF 4 ] (or [bmim][BF 4 ]) absorption heat transformer cycles are studied. • Influence of various operating conditions on cycle’s performance is investigated. • Performance comparisons with H 2 O + LiBr and TFE + TEGDME cycles are done. • Enthalpy data for TFE + [emim][BF 4 ] (or [bmim][BF 4 ]) liquid mixtures are calculated. • TFE + [emim][BF 4 ] (or [bmim][BF 4 ]) cycles have higher gross temperature lift (GTL). - Abstract: A detailed thermodynamic performance analysis of a single-stage absorption heat transformer and double absorption heat transformer cycles using new working pairs composed of ionic liquids (1-ethyl-3-methylimidazolium tetrafluoroborate ([emim][BF 4 ]) and 1-butyl-3-methylimidazolium tetrafluoroborate ([bmim][BF 4 ])) as absorbent and 2,2,2-trifluoroethanol (TFE) as refrigerant has been studied. Several performance indicators were used to evaluate and compare the performance of the cycles using the TFE + [emim][BF 4 ] and TFE + [bmim][BF 4 ] working pairs with the conventional H 2 O + LiBr and organic TFE + TEGDME working pairs. The obtained results show that the ionic liquid based working pairs are suitable candidates to replace the conventional H 2 O + LiBr working pairs in order to avoid the disadvantages associated with it mainly crystallization and corrosion and also they perform better (higher gross temperature lift) than TFE + TEGDME working pair at several operating conditions considered in this work

  13. Method of manufacturing neutron protection materials

    Energy Technology Data Exchange (ETDEWEB)

    Kakibana, Hidetake; Okamoto, Masazane; Fujii, Yasuhiko; Koguchi, Noboru; Takesute, Morito; Miyamatsu, Tokuhisa

    1985-06-22

    To obtain protection materials easily moldable, flexible and capable of minimizing the workers' neutron exposure dose, a fine fiberous assembly is prepared by dispersing compounds of atoms having neutron absorbing performance such as Li or B, for example, finely powderous compounds of LiF or /sup 6/LiF into a solution of spinnable polymer, particularly, polyolefin polymer such as polyethylene in CH/sub 2/Cl and then flash spinning them. The fine fibers are fabricated into mat-like material, blankets, cloths and the likes for use in neutron exposure protection. In the case of neutron irradiation therapy, protection materials of reduced weight, flexible and giving preferred contact with human body can be obtained with ease for protecting the regions other than the lesion area.

  14. Absorbed in the task : Personality measures predict engagement during task performance as tracked by error negativity and asymmetrical frontal activity

    NARCIS (Netherlands)

    Tops, Mattie; Boksem, Maarten A. S.

    2010-01-01

    We hypothesized that interactions between traits and context predict task engagement, as measured by the amplitude of the error-related negativity (ERN), performance, and relative frontal activity asymmetry (RFA). In Study 1, we found that drive for reward, absorption, and constraint independently

  15. A contribution to the development of the modular neutron detector (DEMON): performance evaluation through measurements and simulations

    International Nuclear Information System (INIS)

    Mouatassim, S.

    1994-07-01

    The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON's modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite 126 Ba system formed in the 19 F + 107 Ag entrance channel. (from author) 70 figs., 99 refs

  16. Performance of a neutron transport code with full phase space decomposition on the Cray Research T3D

    International Nuclear Information System (INIS)

    Dorr, M.R.; Salo, E.M.

    1995-01-01

    We present performance results obtained on a 128-node Cray Research T3D computer by a neutron transport code implementing a standard mtiltigroup, discrete ordinates algorithm on a three-dimensional Cartesian grid. After summarizing the implementation strategy used to obtain a full decomposition of phase space (i.e., simultaneous parallelization of the neutron energy, directional and spatial variables), we investigate the scalability of the fundamental source iteration step with respect to each phase space variable. We also describe enhancements that have enabled performance rates approaching 10 gigaflops on the full 128-node machine

  17. Studies performed on neutron-irradiated copper-doped iron specimens by means of neutron small-angle scattering

    International Nuclear Information System (INIS)

    Naraghi, M.

    1978-01-01

    By means of neutron small-angle scattering precipitation arising from heat-treatment and reactor irradiation in copper-alloyed iron specimens were studied. Copper content varried between 0 and 1.5%, irradiation temperature between 310 and 563K. The specimens had been cooled from the melt partly fast, partly slowely. By taking account of magnetic scattering and by investigating the azimuthal dependence of the total scattering it became possible to distinguish between copper precipitations and vacancy agglomerates. The most obvious effect in the slowly cooled specimens after irradiation with 2-10 19 fast neutrons per cm 2 at a temperature of 563 K is the existence of copper agglomerates with diameters of the order of magnitude of 5nm. Precipitation already occurs to a much lesser extent by the influence of temperature alone. Fast cooling from the melt or low irradiation temperature reduce precipitation during reactor irradiation. Moreover, there are indications on the formation of vacancy accumulations and dislocation rings, the latter especially in the fast cooled specimens. (orig.) [de

  18. About the possibility of using the field of the portable neutron generator for treatment of oncological diseases

    International Nuclear Information System (INIS)

    Stoyanov, A.Ph.; Dovbnya, A.N.; Tsymbal, V.A.

    2017-01-01

    The possibility of using a portable neutron generator (PNG) for the treatment of oncological diseases is being considered. It has been shown that when using PNG as a neutron source, it is possible to ensure sufficient therapeutic impact on sick cells, with minimal damage to healthy cells. It's about applying PNG in a brachytherapy tumor. It is important to note that the presence of a narrow ion- pipe- needle allows a neutron source to be placed close to the tumor, and thus to increase therapeutic influence. Numerical estimates of the density of neutrons and the consumed dose when using PNG for brachytherapy performed, it is shown that, for a short period of time (approx 1 minute), sufficient dose of radiation for therapy is absorbed. The calculations of the neutron field and absorbed dose are accomplished through a computer program developed by the authors based on the Monte Carlo method, designed to simulate the generation, movement, braking and absorption of neutrons.

  19. Neutron absorber inserts for 55-gal drums

    International Nuclear Information System (INIS)

    Wilson, R.E.; Kim, Y.S.; Toffer, H.

    2000-01-01

    Transport and temporary storage of more than 200 g of fissile material in 55-gal drums at the Rocky Flats Environmental Technology Site (RFETS) have received significant attention during the cleanup mission. This paper discusses successful applications and results of extensive computer studies. Interim storage and movement of fissile material in excess of standard drum limits (200 g) in a safe configuration have been accomplished using special drum inserts. Such inserts have constrained the contents of a drum to two 4-ell bottles. The content of the bottles was limited to 600 g Pu or U in solution or a total of 1200 g for the entire drum. The inserts were a simple design constructed of stainless steel, forming a vertical cylindrical pipe into which two bottles, one on top of the other, could be centered in the drum. The remaining drum volume was configured to preclude any additional bottle placement external to the vertical cylinder. Such inserts in drums were successfully used in moving high-concentration solution from one building to another for chemical processing. Concern about the knowledge of fissile material concentration in bottles prompted another study for drum inserts. The past practice had been to load up to fourteen 4-ell bottles into 55-gal drums, provided the fissile material concentration was < 6 g fissile/ell, and the total drum contents of 200 g fissile was not exceeded. Only one determination of the solution concentration was needed. An extensive safety analysis concluded that a single measurement of bottle content could not ensure compliance with double-contingency-criterion requirements. A second determination of the bottle contents was required before bottles could be placed in a 55-gal drum. Al alternative to a dual-measurement protocol, which is for bolstering administrative control, was to develop an engineered safety feature that would eliminate expensive tests and administrative decisions. A drum insert design was evaluated that would ensure subcriticality in a drum even if the concentration of fissile material solution exceeded 6 g/ell. The challenge posed to the study was to maximize the number of bottles with a concentration up to 150 g/ell that could be safely placed into a drum. A concentration of 150 g of fissile material per liter is the maximum expected solution concentration in bottles at RFETS. A series of computer analyses were undertaken to design special drum inserts. A design was completed for a 55-gal drum insert to hold up to ten bottles containing 600 g fissile material in solution per bottle or 6 kg per drum. This insert would protect against accidental criticality caused by misloading high-concentration-solution bottles into a drum designated for low-concentration-bottle transport. The drum with the inserts could be used for transport or temporary storage of high-concentration solutions in 4-ell bottles. Conceptually, the drum with its inserts could also be employed for interim storage and transport of fissile material in forms other than solutions in 4-ell bottles

  20. Electrochemical Corrosion Testing of Neutron Absorber Materials

    International Nuclear Information System (INIS)

    Tedd Lister; Ron Mizia; Arnold Erickson; Tammy Trowbridge

    2007-01-01

    This report summarizes the results of crevice-corrosion tests for six alloys in solutions representative of ionic compositions inside the Yucca Mountain waste package should a breech occur. The alloys in these tests are Neutronit A978a (ingot metallurgy, hot rolled), Neutrosorb Plus 304B4 Grade Ab (powder metallurgy, hot rolled), Neutrosorb Plus 304B5 Grade Ab (powder metallurgy, hot rolled), Neutrosorb Plus 304B6 Grade Ab (powder metallurgy, hot rolled), Ni-Cr-Mo-Gd alloy2 (ingot metallurgy, hot rolled), and Alloy 22 (ingot metallurgy, hot rolled)

  1. Slow-Absorbing Modified Starch before and during Prolonged Cycling Increases Fat Oxidation and Gastrointestinal Distress without Changing Performance.

    Science.gov (United States)

    Baur, Daniel A; Vargas, Fernanda de C S; Bach, Christopher W; Garvey, Jordan A; Ormsbee, Michael J

    2016-06-25

    While prior research reported altered fuel utilization stemming from pre-exercise modified starch ingestion, the practical value of this starch for endurance athletes who consume carbohydrates both before and during exercise is yet to be examined. The purpose of this study was to determine the effects of ingesting a hydrothermally-modified starch supplement (HMS) before and during cycling on performance, metabolism, and gastrointestinal comfort. In a crossover design, 10 male cyclists underwent three nutritional interventions: (1) a commercially available sucrose/glucose supplement (G) 30 min before (60 g carbohydrate) and every 15 min during exercise (60 g∙h(-1)); (2) HMS consumed at the same time points before and during exercise in isocaloric amounts to G (Iso HMS); and (3) HMS 30 min before (60 g carbohydrate) and every 60 min during exercise (30 g·h(-1); Low HMS). The exercise protocol (~3 h) consisted of 1 h at 50% Wmax, 8 × 2-min intervals at 80% Wmax, and 10 maximal sprints. There were no differences in sprint performance with Iso HMS vs. G, while both G and Iso HMS likely resulted in small performance enhancements (5.0%; 90% confidence interval = ±5.3% and 4.4%; ±3.2%, respectively) relative to Low HMS. Iso HMS and Low HMS enhanced fat oxidation (31.6%; ±20.1%; very likely (Iso); 20.9%; ±16.1%; likely (Low), and reduced carbohydrate oxidation (-19.2%; ±7.6%; most likely; -22.1%; ±12.9%; very likely) during exercise relative to G. However, nausea was increased during repeated sprints with ingestion of Iso HMS (17 scale units; ±18; likely) and Low HMS (18; ±14; likely) vs. G. Covariate analysis revealed that gastrointestinal distress was associated with reductions in performance with Low HMS vs. G (likely), but this relationship was unclear with Iso HMS vs. G. In conclusion, pre- and during-exercise ingestion of HMS increases fat oxidation relative to G. However, changes do not translate to performance improvements, possibly owing to HMS

  2. Slow-Absorbing Modified Starch before and during Prolonged Cycling Increases Fat Oxidation and Gastrointestinal Distress without Changing Performance

    Directory of Open Access Journals (Sweden)

    Daniel A. Baur

    2016-06-01

    Full Text Available While prior research reported altered fuel utilization stemming from pre-exercise modified starch ingestion, the practical value of this starch for endurance athletes who consume carbohydrates both before and during exercise is yet to be examined. The purpose of this study was to determine the effects of ingesting a hydrothermally-modified starch supplement (HMS before and during cycling on performance, metabolism, and gastrointestinal comfort. In a crossover design, 10 male cyclists underwent three nutritional interventions: (1 a commercially available sucrose/glucose supplement (G 30 min before (60 g carbohydrate and every 15 min during exercise (60 g∙h−1; (2 HMS consumed at the same time points before and during exercise in isocaloric amounts to G (Iso HMS; and (3 HMS 30 min before (60 g carbohydrate and every 60 min during exercise (30 g·h−1; Low HMS. The exercise protocol (~3 h consisted of 1 h at 50% Wmax, 8 × 2-min intervals at 80% Wmax, and 10 maximal sprints. There were no differences in sprint performance with Iso HMS vs. G, while both G and Iso HMS likely resulted in small performance enhancements (5.0%; 90% confidence interval = ±5.3% and 4.4%; ±3.2%, respectively relative to Low HMS. Iso HMS and Low HMS enhanced fat oxidation (31.6%; ±20.1%; very likely (Iso; 20.9%; ±16.1%; likely (Low, and reduced carbohydrate oxidation (−19.2%; ±7.6%; most likely; −22.1%; ±12.9%; very likely during exercise relative to G. However, nausea was increased during repeated sprints with ingestion of Iso HMS (17 scale units; ±18; likely and Low HMS (18; ±14; likely vs. G. Covariate analysis revealed that gastrointestinal distress was associated with reductions in performance with Low HMS vs. G (likely, but this relationship was unclear with Iso HMS vs. G. In conclusion, pre- and during-exercise ingestion of HMS increases fat oxidation relative to G. However, changes do not translate to performance improvements, possibly owing to HMS

  3. GPU-based high performance Monte Carlo simulation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

    2009-07-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  4. GPU-based high performance Monte Carlo simulation in neutron transport

    International Nuclear Information System (INIS)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A.

    2009-01-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  5. Enhanced microwave absorption performance of lightweight absorber based on reduced graphene oxide and Ag-coated hollow glass spheres/epoxy composite

    International Nuclear Information System (INIS)

    Wang, Junpeng; Sun, Yu; Chen, Wei; Wang, Tao; Xu, Renxin; Wang, Jun

    2015-01-01

    Using a combination of Ag-coated hollow glass spheres (HGS@Ag) and a small quantity of graphene sheets within the epoxy matrix, we have prepared a novel lightweight high efficiency microwave absorption composite. Compared with pure HGS@Ag and graphene composite, the −10 dB absorption bandwidth and the minimum reflection loss of the novel composite are improved. Reflection loss exceeding −20 dB is obtained for composites in a wide frequency range and the minimum reflection loss reaches −46 dB while bandwidth less than −10 dB can reach up to 4.1 GHz when an appropriate absorber thickness between 2 and 3.5 mm is chosen. The enhanced microwave absorption performance of the novel composite is due to the enhanced dielectric response, enhanced conductivity, and the trap of electromagnetic radiation with increased propagation paths by multiple reflections

  6. Performance Study of an aSi Flat Panel Detector for Fast Neutron Imaging of Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Schumann, M.; Mauerhofer, E. [Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety, Forschungszentrum Juelich GmbH, 52425 Juelich (Germany); Engels, R.; Kemmerling, G. [Central Institute for Engineering, Electronics and Analytics - Electronic Systems, Forschungszentrum Juelich GmbH, 52425 Juelich (Germany); Frank, M. [MATHCCES - Department of Mathematics, RWTH Aachen University, 52062 Aachen (Germany); Havenith, A.; Kettler, J.; Klapdor-Kleingrothaus, T. [Institute of Nuclear Engineering and Technology Transfer, RWTH Aachen University, 52062 Aachen (Germany); Schitthelm, O. [Corporate Technology, Siemens AG, 91058 Erlangen (Germany)

    2015-07-01

    Radioactive waste must be characterized to check its conformance for intermediate storage and final disposal according to national regulations. For the determination of radio-toxic and chemo-toxic contents of radioactive waste packages non-destructive analytical techniques are preferentially used. Fast neutron imaging is a promising technique to assay large and dense items providing, in complementarity to photon imaging, additional information on the presence of structures in radioactive waste packages. Therefore the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA) using 14 MeV neutrons is studied in a cooperation framework of Forschungszentrum Juelich GmbH, RWTH Aachen University and Siemens AG. However due to the low neutron emission of neutron generators in comparison to research reactors the challenging task resides in the development of an imaging detector with a high efficiency, a low sensitivity to gamma radiation and a resolution sufficient for the purpose. The setup is composed of a commercial D-T neutron generator (Genie16GT, Sodern) with a surrounding shielding made of polyethylene, which acts as a collimator and an amorphous silicon flat panel detector (aSi, 40 x 40 cm{sup 2}, XRD-1642, Perkin Elmer). Neutron detection is achieved using a general propose plastic scintillator (EJ-260, Eljen Technology) linked to the detector. The thermal noise of the photodiodes is reduced by employing an entrance window made of aluminium. Optimal gain and integration time for data acquisition are set by measuring the response of the detector to the radiation of a 500 MBq {sup 241}Am-source. Detector performance was studied by recording neutron radiography images of materials with various, but well known, chemical compositions, densities and dimensions (Al, C, Fe, Pb, W, concrete, polyethylene, 5 x 8 x 10 cm{sup 3}). To simulate gamma-ray emitting waste radiographs in presence of a gamma-ray sources ({sup 60}Co, {sup 137}Cs, {sup 241

  7. Results of the gamma-neutron mapper performance test on 55-gallon drums at the RWMC

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Lawrence, R.S.; Roybal, L.G.; Svoboda, J.M.; Harker, D.J.; Thompson, D.N.; Carpenter, M.V.; Josten, N.E.

    1995-07-01

    The primary purpose of the gamma-neutron mapper (G at sign) is to provide accurate and quantitative spatial information of the gamma-ray and neutron radiation fields as a function of position about the excavation of a radioactive waste site. The GNM is designed to operate remotely and can be delivered to any point on an excavation by the robotic gantry crane developed by the dig-face project at the Idaho National Engineering Laboratory (INEL). It can also be easily adapted to other delivery systems. The GNM can be deployed over a waste site at a predetermined scan rate and has sufficient accuracy to identify and quantify radioactive contaminants of importance. The results reported herein are from a performance test conducted at the Transuranic Storage Area, Building 628, of the Radioactive Waste Management Complex located at the INEL. This building is an active interim-storage area for 55-gal drums of transuranic waste from the Department of Energy's Rocky Flats Plant. The performance test consisted of scanning a stack of drums five high by five wide. Prior to the test, radiation fields were measured by a health physicist at the center of the drums and ranged from 0.5 mR/h to 35 mR/h. Scans of the drums using the GNM were taken at standoff distances from the vertical drum stack of 15 cm, 30 cm, 45 cm, and 90 cm. Data were acquired at scan speeds of 7.5 cm/s and 15 cm/s. The results of these scans and a comparison of these results with the manifests of these drums are compared and discussed

  8. Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

    International Nuclear Information System (INIS)

    Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

    2007-01-01

    Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr

  9. Performance evaluation of a natural-convection solar air-heater with a rectangular-finned absorber plate

    International Nuclear Information System (INIS)

    Fakoor Pakdaman, M.; Lashkari, A.; Basirat Tabrizi, H.; Hosseini, R.

    2011-01-01

    This paper deals with an experimental investigation to evaluate different thermal characteristics of a natural-convection flat-plate solar air-heater with longitudinal rectangular fins array. Having determined the thermal performance of the system a Nusselt number correlation is presented for such finned duct devices. In the presented empirical model which may have industrial applications, solar radiation and ambient temperature have been considered as independent parameters. Other characteristics of the system such as different dimensionless variables, plates and outflow temperatures, efficiency, and mass flow rate have been empirically modeled based on these variables. The particular difference in this study in comparison with the other similar studies is the presentation of an empirical model for rectangular-finned solar air-heaters. This model proposes design concepts and rules of thumb, and demonstrates the calculations of the design parameters. Based on the order of magnitude analysis, solar radiation has been found to be the main parameter which characterizes the thermal behavior of the system. Besides, exergy analysis has been carried out, and optimum conditions in which the system has the highest performance have been determined.

  10. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  11. Analysis of electrophoretic soil humic acids fractions by reversed-phase high performance liquid chromatography with on-line absorbance and fluorescence detection.

    Science.gov (United States)

    Trubetskoj, Oleg A; Richard, Claire; Guyot, Ghislain; Voyard, Guillaume; Trubetskaya, Olga E

    2012-06-22

    A combination of reversed-phase high performance liquid chromatography (RP HPLC) with on-line absorbance and fluorescence detection was used for analysis of chernozem soil humic acids (HAs) and their fractions A, B and C+D with different electrophoretic mobility (EM) and molecular size (MS). Samples were injected onto the column at the identical volume and absorbance. All chromatograms exhibit the resolution of seven peaks. The estimation of relative recovery of HAs and fractions from the reverse-phase column has been done. High MS fraction A, which possesses the low EM, is essentially more hydrophobic (73% of the fraction amount remained adsorbed on the column) and aliphatic than medium MS and EM fraction B (33% of the fraction amount remained adsorbed on the column). The most hydrophilic and aromatic properties belong to low MS fraction C+D, which possess the highest EM and practically was not adsorbed on the column. The hydrophobicity of the bulk HAs lies within the range of fractions hydrophobicity. The absorption spectra of bulk HAs, electrophoretic fractions A, B, C+D and corresponding RP HPLC peaks were featureless but had differences in the values of absorbance ratio at 300 and 400 nm (A3/A4). For fractions A and B this ratio gradually decreased from peak 1 to 7 (from 3.05 to 2.80 and 3.00 to 2.40, respectively). This trend was less pronounced in HAs and practically absent in fraction C+D, where ratio A3/A4 varied within a small range. The strong relationship between fluorescence properties, EM, MS, polarity and aliphaticity/aromaticity of HAs fractions was found. Humic and protein-like fluorescence had different polarity nature. The protein-like fluorescence is located in humic material which irreversibly adsorbed on the reverse-phase column and not subjected to RP HPLC characterization. The humic-like fluorescence at Ex/Em 270/450 nm is mostly located in the hydrophilic peak of low MS fraction C+D. Taking into account that high MS fraction A consisted

  12. The CERN n_TOF Facility: Neutron Beams Performances for Cross Section Measurements

    CERN Document Server

    Chiaveri, E; Andrzejewski, J; Audouin, L; Barbagallo, M; Bécares, V; Bečvář, F; Belloni, F; Berthoumieux, E; Billowes, J; Boccone, V; Bosnar, D; Brugger, M; Calviani, M; Calviño, F; Cano-Ott, D; Carrapiço, C; Cerutti, F; Chin, M; Colonna, N; Cortés, G; Cortés-Giraldo, M A; Diakaki, M; Domingo-Pardo, C; Duran, I; Dressler, R; Dzysiuk, N; Eleftheriadis, C; Ferrari, A; Fraval, K; Ganesan, S; García, A R; Giubrone, G; Gómez-Hornillos, M B; Gonçalves, I F; González-Romero, E; Griesmayer, E; Guerrero, C; Gunsing, F; Gurusamy, P; Hernández-Prieto, A; Jenkins, D G; Jericha, E; Kadi, Y; Käppeler, F; Karadimos, D; Kivel, N; Koehler, P; Kokkoris, M; Krtička, M; Kroll, J; Lampoudis, C; Langer, C; Leal-Cidoncha, E; Lederer, C; Leeb, H; Leong, L S; Losito, R; Mallick, A; Manousos, A; Marganiec, J; Martínez, T; Massimi, C; Mastinu, P F; Mastromarco, M; Meaze, M; Mendoza, E; Mengoni, A; Milazzo, P M; Mingrone, F; Mirea, M; Mondalaers, W; Paradela, C; Pavlik, A; Perkowski, J; Plompen, A; Praena, J; Quesada, J M; Rauscher, T; Reifarth, R; Riego, A; Robles, M S; Roman, F; Rubbia, C; Sabaté-Gilarte, M; Sarmento, R; Saxena, A; Schillebeeckx, P; Schmidt, S; Schumann, D; Tagliente, G; Tain, J L; Tarrío, D; Tassan-Got, L; Tsinganis, A; Valenta, S; Vannini, G; Variale, V; Vaz, P; Ventura, A; Versaci, R; Vermeulen, M J; Vlachoudis, V; Vlastou, R; Wallner, A; Ware, T; Weigand, M; Weiss, C; Wright, T; Žugec, P

    2014-01-01

    This paper presents the characteristics of the existing CERN n\\_TOF neutron beam facility (n\\_TOF-EAR1 with a flight path of 185 meters) and the future one (n\\_TOF EAR-2 with a flight path of 19 meters), which will operate in parallel from Summer 2014. The new neutron beam will provide a 25 times higher neutron flux delivered in 10 times shorter neutron pulses, thus offering more powerful capabilities for measuring small mass, low cross section and/or high activity samples.

  13. Research on preparation and performance of graphite cement-based materials used for fast neutron shielding

    International Nuclear Information System (INIS)

    Xu Jun; Kang Qing; Shen Zhiqiang; Wang Zhenggang; Wang Zhiqiang

    2014-01-01

    Measurements have been carried out to investigate the 14.8 MeV neutron attenuation properties for 3 kinds of cement-graphite composites. In comparison with the void group, the 14.8 MeV neutron attenuation properties of cement-graphite composites raised not clearly in 8 mm thickness, and drop not remarkably in 40 mm thickness; with the increase of graphite content and the thickness, the 14.8 MeV neutron attenuation properties were enhanced clearly. The data may be useful to the radiation shielding design of neutron. (authors)

  14. Performance evaluation of the QC-6PLUS quality control system in terms of photons and electrons absorbed doses to water

    International Nuclear Information System (INIS)

    Teixeira, Flavia Cristina da Silva

    2004-06-01

    The quality of the treatment in radiotherapy depends on the necessary knowledge of the liberated dose in the tumor and of several other physical parameters and dosimetric that characterize the profile of the radiation field. Worrying about the reliability of some commercial equipment that aim at determining the main parameters of a radiation field in a practical way for daily checks in an institution with radiotherapy service, in this work a study of the performance of the quality assurance system, QC6-Plus manufactured by PTW-Freiburg for daily checks, was developed, in order to assure the use of this equipment with larger reliability level in the routine of quality assurance of the hospitals as well as to make possible its use in the Program of Regulatory Inspections of the Services of Radiotherapy of the Country accomplished by IRD/CNEN. The found results indicate that the system QC6-Plus is perfectly adapted and practical for relative measures of daily and weekly control of the main parameters of clinical beans in agreement with reference values recommended in TECDOC 1151. However for measurements of absolute dose it should not be used because, for beams of electrons the system does not present the necessary characteristics to execute this measure type in agreement with the reference protocol, TRS 398, and for photons of energy 15 MV presented a deviation in relation to the conventional method of absolute dosimetry of 7,7%, that it is a lot above the expected in agreement with TRS 398. (author)

  15. Elements and process for recording direct image neutron radiographs

    International Nuclear Information System (INIS)

    Poignant, R.V. Jr.; Przybylowicz, E.P.

    1975-01-01

    An element is provided for recording a direct image neutron radiograph, thus eliminating the need for a transfer step (i.e., the use of a transfer screen). The element is capable of holding an electrostatic charge and comprises a first layer for absorbing neutrons and generating a current by dissipation of said electrostatic charge in proportion to the number of neutrons absorbed, and a second layer for conducting the current generated by the absorbed neutrons, said neutron absorbing layer comprising an insulative layer comprising neutron absorbing agents in a concentration of at least 10 17 atoms per cm 3 . An element for enhancing the effect of the neutron beam by utilizing the secondary emanations of neutron absorbing materials is also disclosed along with a process for using the device. (U.S.)

  16. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in

  17. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  18. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  19. Nuclear characteristics of epoxy resin as a space environment neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School

    2017-03-15

    In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.

  20. Influence of Fe{sub 3}O{sub 4}/Fe-phthalocyanine decorated graphene oxide on the microwave absorbing performance

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jingwei; Wei, Junji; Pu, Zejun; Xu, Mingzhen; Jia, Kun, E-mail: jiakun@uestc.edu.cn; Liu, Xiaobo, E-mail: liuxb@uestc.edu.cn

    2016-02-01

    Novel graphene oxide@Fe{sub 3}O{sub 4}/iron phthalocyanine (GO@Fe{sub 3}O{sub 4}/FePc) hybrid materials were prepared through a facile one-step solvothermal method with graphene oxide (GO) sheets as template in ethylene glycol. The morphology and structure of the hybrid materials were characterized by scanning electron microscopy (SEM), transmission electron microscopy (TEM), Fourier transform infrared spectrophotometer (FTIR) and X-ray diffraction (XRD), respectively. The results indicated that the monodispersed Fe{sub 3}O{sub 4}/FePc hybrid microspheres were uniformly self-assembled along the surface of GO sheets through electrostatic attraction and the morphology can be tuned by controlling the amount of 4,4′-bis(3,4-dicyanophenoxy)biphenyl (BPH). As the BPH content increases, magnetization measurement of the GO@Fe{sub 3}O{sub 4}/FePc hybrid materials showed that the coercivity increased, while saturation magnetizations decreased. Electromagnetic properties of the hybrid materials were measured in the range of 0.5–18.0 GHz. The microwave absorbing performance enhanced with the increase of BPH content and a maximum reflection loss of −27.92 dB was obtained at 10.8 GHz when the matching thickness was 2.5 mm. Therefore, the novel electromagnetic hybrid materials can be considered as potential materials in the microwave absorbing field. - Highlights: • Graphene oxide was employed to support Fe{sub 3}O{sub 4}/iron phthalocyanine hybrid particles. • The morphology and magnetic properties of obtained particles can be readily tuned. • A maximum microwave reflection loss of −27.92 dB was obtained at 10.8 GHz.

  1. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

  2. Dual-sided microstructured semiconductor neutron detectors (DSMSNDs)

    International Nuclear Information System (INIS)

    Fronk, Ryan G.; Bellinger, Steven L.; Henson, Luke C.; Ochs, Taylor R.; Smith, Colten T.; Kenneth Shultis, J.; McGregor, Douglas S.

    2015-01-01

    Microstructured semiconductor neutron detectors (MSNDs) have in recent years received much interest as high-efficiency replacements for thin-film-coated thermal neutron detectors. The basic device structure of the MSND involves micro-sized trenches that are etched into a vertically-oriented pvn-junction diode that are backfilled with a neutron converting material. Neutrons absorbed within the converting material induce fission of the parent nucleus, producing a pair of energetic charged-particle reaction products that can be counted by the diode. The MSND deep-etched microstructures produce good neutron-absorption and reaction-product counting efficiencies, offering a 10× improvement in intrinsic thermal neutron detection efficiency over thin-film-coated devices. Performance of present-day MSNDs are nearing theoretical limits; streaming paths between the conversion-material backfilled trenches, allow a considerable fraction of neutrons to pass undetected through the device. Dual-sided microstructured semiconductor neutron detectors (DSMSNDs) have been developed that utilize a complementary second set of trenches on the back-side of the device to count streaming neutrons. DSMSND devices are theoretically capable of greater than 80% intrinsic thermal neutron detection efficiency for a 1-mm thick device. The first such prototype DSMSNDs, presented here, have achieved 29.48±0.29% nearly 2× better than MSNDs with similar microstructure dimensions.

  3. Preliminary study on zinc-carbon battery performance by using neutron tomography

    International Nuclear Information System (INIS)

    Abdul Aziz Mohamed; Nor Abidin Ashari; Mohd Zaid Abdullah; Junita Mohamad Saleh; Azraf Azman; Megat Harun AlRashid Megat Ahmad; Rafhayudi Jamro

    2008-08-01

    This paper describes on the discharging characteristic of zinc-carbon batteries (dry cells) by using a neutron imaging technique called a monochromatic neutron tomography. Experiment was conducted on the Nuclear Malaysia neutron tomography prototype instrument which based on 1-dimensional position sensitive neutron detector. The instrument is constructed at the small angle neutron scattering (SANS) beam line built at the one of the beam ports of TRIGA MARK II Research reactor, Malaysian Nuclear Agency, Bangi, Selangor. The main aim of this preliminary experiment was to test the instrument capability on a real industrial component. It was also aimed to understand structural and chemical changes of these battery particles after experiencing a discharging process. In this preliminary work, new and used batteries used were the products of Eveready company. (Author)

  4. Imaging with cold neutrons

    International Nuclear Information System (INIS)

    Lehmann, E.H.; Kaestner, A.; Josic, L.; Hartmann, S.; Mannes, D.

    2011-01-01

    Neutrons for imaging purposes are provided mainly from thermal beam lines at suitable facilities around the world. The access to cold neutrons is presently limited to very few places only. However, many challenging options for imaging with cold neutrons have been found out, given by the interaction behavior of the observed materials with neutrons in the cold energy range (3-10 A). For absorbing materials, the interaction probability increases proportionally with the wavelength with the consequence of more contrast but less transmission with cold neutrons. Many materials are predominantly scattering neutrons, in particular most of crystalline structural materials. In these cases, cold neutrons play an important role by covering the energy range of the most important Bragg edges given by the lattice planes of the crystallites. This particular behavior can be used for at least two important aspects-choosing the right energy of the initial beam enables to have a material more or less transparent, and a direct macroscopic visualization of the crystalline structure and its change in a manufacturing process. Since 2006, PSI operates its second beam line for neutron imaging, where cold neutrons are provided from a liquid deuterium cold source (operated at 25 K). It has been designed to cover the most current aspects in neutron imaging research with the help of high flexibility. This has been done with changeable inlet apertures, a turbine based velocity selector, two beam positions and variable detector systems, satisfying the demands of the individual investigation. The most important detection system was found to be a micro-tomography system that enables studies in the presently best spatial resolution. In this case, the high contrast from the sample interaction process and the high detection probability for the cold neutrons combines in an ideal combination for the best possible performance. Recently, it was found out that the energy selective studies might become a

  5. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, Pavel, E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Fischer, Ulrich [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, {sup 6}Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  6. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    International Nuclear Information System (INIS)

    Pereslavtsev, Pavel; Bachmann, Christian; Fischer, Ulrich

    2016-01-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, "6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  7. Performance of Large Neutron Detectors Containing Lithium-Gadolinium-Borate Scintillator

    Energy Technology Data Exchange (ETDEWEB)

    Slaughter, David M.; Stuart, Cory R.; Klaass, R. Fred; Merrill, David B. [MSI/Photogenics Division, Orem, Utah (United States)

    2015-07-01

    This paper describes the development and testing of a neutron counter, spectrometer, and dosimeter that is compact, efficient, and accurate. A self-contained neutron detection instrument has wide applications in health physics, scientific research, and programs to detect, monitor, and control strategic nuclear materials (SNM). The 1.3 liter detector head for this instrument is a composite detector with an organic scintillator containing uniformly distributed {sup 6}Li{sub 6}{sup nat}Gd{sup 10}B{sub 3}O{sub 9}:Ce (LGB:Ce) microcrystals. The plastic scintillator acts to slow impinging neutrons and emits light proportional to the energy lost by the neutrons as they moderate in the detector body. Moderating neutrons that have slowed sufficiently capture in one of the Lithium-6, Boron-10, or Gadolinium-157 atoms in the LGB:Ce scintillator, which then releases the capture energy in a characteristic cerium emission pulse. The measured captured pulses indicate the presence of neutrons. When a scintillating fluor is present in the plastic, the light pulse resulting from the neutron moderating in the plastic is paired with the LGB:Ce capture pulse to identify the energy of the neutron. About 2% of the impinging neutrons lose all of their energy in a single collision with the detector. There is a linear relationship between the pulse areas of this group of neutrons and energy. The other 98% of neutrons have a wide range of collision histories within the detector body. When these neutrons are 'binned' into energy groups, each group contains a distribution of pulse areas. This data was used to assist in the unfolding of the neutron spectra. The unfolded spectra were then validated with known spectra, at both neutron emitting isotopes and fission/accelerator facilities. Having validated spectra, the dose equivalent and dose rate are determined by applying standard, regulatory damage coefficients to the measured neutron counts for each energy bin of the spectra. Testing

  8. Parallel microscope-based fluorescence, absorbance and time-of-flight mass spectrometry detection for high performance liquid chromatography and determination of glucosamine in urine.

    Science.gov (United States)

    Xiong, Bo; Wang, Ling-Ling; Li, Qiong; Nie, Yu-Ting; Cheng, Shuang-Shuang; Zhang, Hui; Sun, Ren-Qiang; Wang, Yu-Jiao; Zhou, Hong-Bin

    2015-11-01

    A parallel microscope-based laser-induced fluorescence (LIF), ultraviolet-visible absorbance (UV) and time-of-flight mass spectrometry (TOF-MS) detection for high performance liquid chromatography (HPLC) was achieved and used to determine glucosamine in urines. First, a reliable and convenient LIF detection was developed based on an inverted microscope and corresponding modulations. Parallel HPLC-LIF/UV/TOF-MS detection was developed by the combination of preceding Microscope-based LIF detection and HPLC coupled with UV and TOF-MS. The proposed setup, due to its parallel scheme, was free of the influence from photo bleaching in LIF detection. Rhodamine B, glutamic acid and glucosamine have been determined to evaluate its performance. Moreover, the proposed strategy was used to determine the glucosamine in urines, and subsequent results suggested that glucosamine, which was widely used in the prevention of the bone arthritis, was metabolized to urines within 4h. Furthermore, its concentration in urines decreased to 5.4mM at 12h. Efficient glucosamine detection was achieved based on a sensitive quantification (LIF), a universal detection (UV) and structural characterizations (TOF-MS). This application indicated that the proposed strategy was sensitive, universal and versatile, and it was capable of improved analysis, especially for analytes with low concentrations in complex samples, compared with conventional HPLC-UV/TOF-MS. Copyright © 2015 Elsevier B.V. All rights reserved.

  9. Performance Study of a Cylindrical Parabolic Concentrating Solar Water Heater with Nail Type Twisted Tape Inserts in the Copper Absorber Tube

    Directory of Open Access Journals (Sweden)

    Amit K. Bhakta

    2018-01-01

    Full Text Available This paper reports the overall thermal performance of a cylindrical parabolic concentrating solar water heater (CPCSWH with inserting nail type twisted tape (NTT in the copper absorber tube for the nail twist pitch ratios, 4.787, 6.914 and 9.042, respectively. The experiments are conducted for a constant volumetric water flow rate and during the time period 9:00 a.m. to 15:00 p.m. The useful heat gain, hourly solar energy collected and hourly solar energy stored in this solar water heater were found to be higher for the nail twist pitch ratio 4.787. The above said parameters were found to be at a peak at noon and observed to follow the path of variation of solar intensity. At the start of the experiment, the value of charging efficiency was observed to be maximum, whereas the maximum values of instantaneous efficiency and overall thermal efficiency were observed at noon. The key finding is that the nail twist pitch ratio enhances the overall thermal performance of the CPCSWH.

  10. Adaptive inertial shock-absorber

    International Nuclear Information System (INIS)

    Faraj, Rami; Holnicki-Szulc, Jan; Knap, Lech; Seńko, Jarosław

    2016-01-01

    This paper introduces and discusses a new concept of impact absorption by means of impact energy management and storage in dedicated rotating inertial discs. The effectiveness of the concept is demonstrated in a selected case-study involving spinning management, a recently developed novel impact-absorber. A specific control technique performed on this device is demonstrated to be the main source of significant improvement in the overall efficiency of impact damping process. The influence of various parameters on the performance of the shock-absorber is investigated. Design and manufacturing challenges and directions of further research are formulated. (paper)

  11. Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

    International Nuclear Information System (INIS)

    Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

    2006-01-01

    The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

  12. Absorbing Property of Multi-layered Short Carbon Fiber Absorbing Coating

    OpenAIRE

    Liu, Zhaohui; Tao, Rui; Ban, Guodong; Luo, Ping

    2018-01-01

    The radar absorbing coating was prepared with short carbon fiber asabsorbent and waterborne polyurethane (WPU) as matrix resin. The coating’s absorbing property was tested with vectornetwork analyzer, using aramid honeycomb as air layer which was matched withcarbon fiber coating. The results demonstrate that the single-layered carbonfiber absorbing coating presented relatively poor absorbing property when thelayer was thin, and the performance was slightly improved after the matched airlayer ...

  13. Experiments on the Biological Action of Neutrons Performed in the Former Soviet Union: A Historical Review

    Science.gov (United States)

    2006-10-01

    Exposed to x-rays and Fission Neutrons at Different Stages of Postnatal Development. Radio- biologia , 1982, vol. 22, no. 1, pp. 76-81. 10. Bogatyrev A.V...Cyclotron U-120. Radio- biologia , vol. 24, no. 4, pp. 567-569, 1986. 17. Dokshina G.A., Naumenko L.A., Effect of Mixed Gamma-Neutron Radia- tion on...in Drosophila Spermatogenesis on Dose from Fast Neutrons. Radio- biologia , v. 2, no. 3, pp. 420-421, 1963. 4. Alexandrov I.D., Different RBE for Gene

  14. Simulations and measurements of the performance of a channeled neutron guide for a time-of-flight spectrometer at the NIST Center for Neutron Research

    International Nuclear Information System (INIS)

    Cook, Jeremy C.; Copley, John R.D.

    2004-01-01

    We describe the identification and analysis of the principal sources of intensity loss within the five-channeled neutron guide tube that was originally installed in the chopper section of the Disk Chopper Spectrometer at the National Institute of Standards and Technology Center for Neutron Research. (The purpose of the five channels was to optimize intensity and resolution in three different modes of operation known as ''resolution modes.'') By combining measurements, Monte Carlo simulations, and analytical calculations, we have developed a model that successfully explains performance losses in the original guide. We have used this model to quantify expected returns in performance using a replacement guide in which the principal contributions to the intensity loss are reduced to the minimum achievable with current technology. We have also estimated the intensity gains that would be achieved if one of the limited number of options were adopted for modifying the original guide in a manner likely to produce such gains. We describe factors that affect the performance of the original guide and compare the measured and predicted performance of the modified guide against predictions for the optimal replacement guide. The simulations indicate that the modified guide (which has three channels rather than the original five) produces greater intensity gains over a large incident wavelength band for the low and medium resolution modes, whereas a high quality replacement guide greatly improves performance in the high resolution mode of operation. Because the low and medium resolution modes are most heavily demanded, we opted to modify the guide rather than replace it. We describe the nature of this modification and present intensity measurements that meet or exceed predictions in all resolution modes with no detectable change in the energy resolution nor increase in the instrumental background

  15. Fatigue performance of copper and copper alloys before and after irradiation with fission neutrons

    International Nuclear Information System (INIS)

    Singh, B.N.; Toft, P.; Stubbins, J.F.

    1997-05-01

    The fatigue performance of pure copper of the oxygen free, high conductivity (OFHC) grade and two copper alloys (CuCrZr and CuAl-25) was investigated. Mechanical testing and microstructural analysis were carried out to establish the fatigue life of these materials in the unirradiated and irradiated states. The present report provides the first information on the ability of these copper alloys to perform under cyclic loading conditions when they have undergone significant irradiation exposure. Fatigue specimens of OFHC-Cu, CuCrZr and CuAl-25 were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of ∼2.5 x 10 17 n/m 2 s (E > 1 MeV) to fluence levels of 1.5 - 2.5 x 10 24 n/m 2 s (E > 1 MeV) at ∼47 and 100 deg. C. Specimens irradiated at 47 deg. C were fatigue tested at 22 deg. C, whereas those irradiated at 100 deg. C were tested at the irradiation temperature. The major conclusion of the present work is that although irradiation causes significant hardening of copper and copper alloys, it does not appear to be a problem for the fatigue life of these materials. In fact, the present experimental results clearly demonstrate that the fatigue performance of the irradiated CuAl-25 alloy is considerably better in the irradiated than that in the unirradiated state tested both at 22 and 100 deg. C. This improvement, however, is not so significant in the case of the irradiated OFHC-copper and CuCrZr alloy tested at 22 deg. C. These conclusions are supported by the microstructural observations and cyclic hardening experiments. (au) 4 tabs., 26 ills., 10 refs

  16. Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Malatya Univ. (Turkey). Physics Department

    2015-03-15

    In this study, salt-heavy metal mixtures consisting of 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% UO{sub 2}, 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% NpO{sub 2}, and 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion-fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio (TBR), energy multiplication factor (M), and heat deposition rate, of the fusion-fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

  17. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in neutron transport problems, the authors briefly describe the work done in order to get a vector code. Vectorization principles are presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples are presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example they propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion they prove that Monte Carlo algorithms are very well suited to future vector and parallel computers

  18. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)

  19. Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method

    International Nuclear Information System (INIS)

    Guenay, Mehtap

    2015-01-01

    In this study, salt-heavy metal mixtures consisting of 93-85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2-10% UO 2 , 93-85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2-10% NpO 2 , and 93-85% Li 20 Sn 80 + 5% SFG-PuO 2 and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion-fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio (TBR), energy multiplication factor (M), and heat deposition rate, of the fusion-fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

  20. A low noise ASIC for two dimensional neutron gas detector with performance of high spatial resolution (Contract research)

    International Nuclear Information System (INIS)

    Yamagishi, Hideshi; Toh, Kentaro; Nakamura, Tatsuya; Sakasai, Kaoru; Soyama, Kazuhiko

    2012-02-01

    An ASD-ASIC (Amplifier-Shaper-Discriminator ASIC) with fast response and low noise performances has been designed for two-dimensional position sensitive neutron gas detectors (InSPaD). The InSPaD is a 2D neutron detector system with 3 He gas and provides a high spatial resolution by making distinction between proton and triton particles generated in the gas chamber. The new ASD-ASIC is required to have very low noise, a wide dynamic range, good output linearity and high counting rate. The new ASD-ASIC has been designed by using CMOS and consisted of 64-channel ASDs, a 16-channel multiplexer with LVTTL drivers and sum amplifier system for summing all analog signals. The performances were evaluated by the Spice simulation. It was confirmed that the new ASD-ASIC had very low noise performance, wide dynamic range and fast signal processing functions. (author)

  1. Fe-Substitution for Ni in Misch Metal-Based Superlattice Hydrogen Absorbing Alloys—Part 2. Ni/MH Battery Performance and Failure Mechanisms

    Directory of Open Access Journals (Sweden)

    Tiejun Meng

    2017-09-01

    Full Text Available The electrochemical performance and failure mechanisms of Ni/MH batteries made with a series of the Fe-substituted A2B7 superlattice alloys as the negative electrodes were investigated. The incorporation of Fe does not lead to improved cell capacity or cycle life at either room or low temperature, although Fe promotes the formation of a favorable Ce2Ni7 phase. Fe-substitution was found to inhibit leaching of Al from the metal hydride negative electrode and promote leaching of Co, which could potentially extend the cycle life of the positive electrode. The failure mechanisms of the cycled cells with the Fe-substituted superlattice hydrogen absorbing alloys were analyzed by scanning electron microscopy, energy dispersive spectroscopy and inductively coupled plasma analysis. The failure of cells with Fe-free and low Fe-content alloys is mainly attributed to the pulverization of the metal hydride alloy. Meanwhile, severe oxidation/corrosion of the negative electrode is observed for cells with high Fe-content alloys, resulting in increased internal cell resistance, formation of micro-shortages in the separator and eventual cell failure.

  2. Radiolabeling of intact dosage forms by neutron activation: effects on in vitro performance

    International Nuclear Information System (INIS)

    Parr, A.; Jay, M.

    1987-01-01

    Compressed tablets containing various quantities of stable isotopes of Ba, Er, and Sm for use in neutron activation studies were evaluated for the effect of stable isotope incorporation on tablet hardness and disintegration times. At concentrations likely to be used in scintigraphic studies employing neutron activation as a radiolabeling method, no significant effect on in vitro parameters were observed. While the incorporation of stable isotopes influenced tablet hardness to a greater degree than disintegration time, irradiation of tablets in a neutron flux of 4.4 x 10(13) n/cm2 sec had a direct effect on tablet disintegration time. Thus, future neutron activation studies should focus on minimizing the amount of stable isotope to be incorporated with the formulation while using the shortest feasible irradiation time

  3. Preliminary study of performance of peripheral absorbable magnesium alloy stent%外周动脉可吸收镁合金支架性能初探

    Institute of Scientific and Technical Information of China (English)

    肖耀文; 陈忠; 杨耀国; 寇镭; 郑丰; 张炳春; 白树功; 郑焕勤; 刘博文

    2016-01-01

    Objective To investigate the performance and safety of domestic peripherally absorbable magnesium alloy stent.Methods Sixteen absorbable magnesium alloy stents were implanted in the bilateral iliac-femoral arteries of 8 Chinese Agricultural University minipigs selected randomly with computer.The mean weight was (21.19± 1.93)kg.One pig was randomly selected by computer on 30th,45th,60th,75th,90th,105th,120th and 135th days after stent implantation and both stents were removed.The biochemical,imaging and histopathology index of the pig before and after stent placement were analyzed,t test was used to compare data collected before and after stenting.Results The biochemical index changes showed no statistical significance after stent implantation,the DSA indicated that vessel diameter was (4.39± 0.26) mm before stent implantation and (3.85±0.71)mm on the 30th to 135th day afterimplatation (t=3.041,P=0.008),the balance of vessel diameter before and after stenting ranged from 0 to 2.29 mm,with minimun on 30th day,and maximum on 90th day;HE and verhoeffvangieson showed significant stenosis of the arteries occurred on 60'h,90th and 120th day.From the 30th day to 135th day after stenting,the average diameter of residual lumen was (1.38±0.18) mm,the average area of residual lumen was (1.51±0.41) mm2,the average area of internal elastic lamina was (1.91t0.22) mm2,the average area of external elastic lamina was (4.79± 0.79) mm2,the average thickness of endothelium was (0.11 ±0.06) mm,the average area of new intima was (0.40±0.29) mm2,the average narrow area percentage was (21.96± 16.48)%.The stent absorbed gradually,and not completely absorbed until 135th day after stenting.Endometrial cells completely covered stent surface,and the stent did not break or collapse at all.Conclusions The peripherally absorbable magnesium alloy stents is safe and effective in animal experiment.Application of this stent in human should be investigated with further studies.%目的 初

  4. Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

    International Nuclear Information System (INIS)

    Raisic, N.; Klinc, T.

    1968-11-01

    The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

  5. The Clatterbridge high-energy neutron therapy facility: specification and performance

    International Nuclear Information System (INIS)

    Bonnett, D.E.; Blake, S.W.; Shaw, J.E.; Bewley, D.K.

    1988-01-01

    A high energy neutron therapy facility has been installed at the Douglas Cyclotron Centre, Clatterbridge Hospital Merseyside, to extend M.R.C. clinical trials of fast neutrons. The neutron beam is produced by bombarding a beryllium target with 62 MeV protons. The target is isocentrically mounted with potential for 360 0 rotation, with a fully variable collimator, giving a range of rectilinear field sizes from 5 cm x 5 cm to 30 cm x 30 cm. Basic neutron beam data including output, field flatness, penumbra and depth-dose data have been measured. For a 10 cm x 10 cm field, 50% depth dose occurs at 16.2 cm in water and output is 1.63 cGy μ A -1 min -1 at maximum dose depth. Effectiveness of the target shielding and neutron-induced radioactivity in the treatment head were also measured. It is concluded that the equipment meets design specifications and fully satisfies criticisms of earlier neutron therapy equipment. A full radiation survey showed that radiation levels present no significant staff hazard. (UK)

  6. Metamaterial electromagnetic wave absorbers.

    Science.gov (United States)

    Watts, Claire M; Liu, Xianliang; Padilla, Willie J

    2012-06-19

    The advent of negative index materials has spawned extensive research into metamaterials over the past decade. Metamaterials are attractive not only for their exotic electromagnetic properties, but also their promise for applications. A particular branch-the metamaterial perfect absorber (MPA)-has garnered interest due to the fact that it can achieve unity absorptivity of electromagnetic waves. Since its first experimental demonstration in 2008, the MPA has progressed significantly with designs shown across the electromagnetic spectrum, from microwave to optical. In this Progress Report we give an overview of the field and discuss a selection of examples and related applications. The ability of the MPA to exhibit extreme performance flexibility will be discussed and the theory underlying their operation and limitations will be established. Insight is given into what we can expect from this rapidly expanding field and future challenges will be addressed. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  7. Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ulrich, Timothy J. II [Los Alamos National Laboratory; Lafleur, Adrienne M. [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Los Alamos National Laboratory; Bolind, Alan M. [Los Alamos National Laboratory

    2012-07-16

    An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  8. Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Krstic, D.; Markovic, V.M.; Jovanovic, Z.; Milenkovic, B.; Nikezic, D.; Atanackovic, J.

    2014-01-01

    Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. The difference in evaluated dose in cancer and normal lung tissue suggests that BNCT could be applied for the treatment of cancers. The difference in exposure of cancer and healthy tissue can be observed, so the healthy tissue can be spared from damage. An absorbed dose ratio of metastatic tissue-to-the healthy tissue was ∼5. Absorbed dose to all other organs was low when compared with the lung dose. Absorbed dose depth distribution shows that BNC therapy can be very useful in the treatments for tumour. The ratio of the tumour absorbed dose and irradiated healthy tissue absorbed dose was also ∼5. It was seen that an elliptical neutron field was better irradiation choice. (authors)

  9. Prototype Stilbene Neutron Collar

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shumaker, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Verbeke, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation of list mode neutron correlation data on various fuel assemblies including some with neutron absorbers (burnable Gd poisons). Calibration curves (doubles vs 235U linear mass density) are presented for both thermal and fast (with Cd lining) modes of operation. It is shown that the stilbene collar meets or exceeds the current capabilities of He-3 based neutron collars. A self-consistent assay methodology, uniquely suited to the stilbene collar, using triples is described which complements traditional assay based on doubles calibration curves.

  10. Use of helium-neon laser for the prevention of acute radiation reaction of the skin in neutron-beam therapy of head and neck tumors

    International Nuclear Information System (INIS)

    Popovich, V.I.; Musabaeva, L.I.; Kitsmanyuk, Z.D.; Lavrenkov, K.A.

    1991-01-01

    Preliminary data on helium-neon laser usage to prevent acute radiation skinresponse in patients with head and neck neoplasm were presented in case of fast neutrons therapy with average energy of ≅ 6.3 MeV. Irradiation was performed by 2 fractions a week with single absorbed focal dose of 1.2-1.4 Gy and the dose for the skin was 2-2.2 Gy. RBE of the fast neutrons comprised ∼ 3. Some patients were subjected to neutron therapy in combination with helium-neon laser treatment, the others underwent only neutron therapy. Combination of neutron and helium-neon laser therapy increased skin resistance to neutron irradiation. Combined treatment with neutrons and helium-neon laser decreased development of humid epidermitis by half than in case of neutron treatment alone

  11. Neutron stars

    International Nuclear Information System (INIS)

    Irvine, J.M.

    1978-01-01

    The subject is covered in chapters entitled: introduction (resume of stellar evolution, gross characteristics of neutron stars); pulsars (pulsar characteristics, pulsars as neutron stars); neutron star temperatures (neutron star cooling, superfluidity and superconductivity in neutron stars); the exterior of neutron stars (the magnetosphere, the neutron star 'atmosphere', pulses); neutron star structure; neutron star equations of state. (U.K.)

  12. Burnable absorber for the PIK reactor

    International Nuclear Information System (INIS)

    Gostev, V.V.; Smolskii, S.L.; Tchmshkyan, D.V.; Zakharov, A.S.; Zvezdkin, V.S.; Konoplev, K.A.

    1998-01-01

    In the reactor PIK design a burnable absorber is not used and the cycle duration is limited by the rods weight. Designed cycle time is two weeks and seams to be not enough for the 100 MW power research reactor equipped by many neutron beams and experimental facilities. Relatively frequent reloading reduces the reactor time on full power and in this way increases the maintenance expenses. In the reactor core fuel elements well mastered by practice are used and its modification was not approved. We try to find the possibilities of installation in the core separate burnable elements to avoid poison of the fuel. It is possible to replace a part of the fuel elements by absorbers, since the fuel elements are relatively small (diameter 5.15mm, uranium 235 content 7.14g) and there are more then 3800 elements in the core. Nevertheless, replacing decreases the fuel burnup and its consumption. In the PIK fuel assembles a little part of the volume is occupied by the dumb elements to create a complete package of the assembles shroud, that is necessary in the hydraulic reasons. In the presented report the assessment of such a replacement is done. As a burnable material Gadolinium was selected. The measurements or the beginning of cycle were performed on the critical facility PIK. The burning calculation was confirmed by measurements on the 18MW reactor WWR-M. The results give the opportunity to twice the cycle duration. The proposed modification of the fuel assembles does not lead to alteration in the other reactor systems, but it touch the burned fuel reprocessing technology. (author)

  13. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

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

    1993-01-01

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  14. Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-372

    Energy Technology Data Exchange (ETDEWEB)

    Fanchini, Erica [INFN/ANN and SCINTILLA groups, Isituto Nazionale di Fisica Nucleare - INFN (Italy)

    2015-07-01

    A Radiation Portal Monitor (RPM) was developed by the Istituto Nazionale di Fisica Nucleare (INFN) and Ansaldo Nucleare (ANN) within the FP7 SCINTILLA European project. The system was designed to detect both gamma and neutron radiation with a single technology. It is conceived to monitor vehicle and cargo containers in transits across borders or ports, to find radioactive elements and to avoid illegal trafficking of strategic nuclear materials. The system is based on a {sup 3}He-free neutron detection technology using plastic scintillators coupled to Gadolinium to detect and discriminate gamma from neutron signals. During the 3 years of the SCINTILLA project the construction and test of the first two prototypes drove the definition of the final layout of a full RPM system consisting of two twin pillars as a portal for vehicle and cargo container scan. A custom System Control Software (SCS) manages the electronics of the RPM, the ancillary devices and the data analysis. The combination of the detector layout and of the software functionalities enables both to distinguish neutrons and gammas and to identify the energy range of a detected gamma source. The system was initially characterized via static tests with gamma and neutron sources in the INFN laboratory. These measurements were used to calibrate the detector, evaluate the response of the single pillars as well as of the full system, and optimize the RPM configuration and discrimination algorithm. During this phase, specific tests were performed to study the stability over time of the system, monitoring the measured the neutron and gamma count rates over periods of several weeks. The results allow us to demonstrate the reliability and robustness of the RPM. In a second time the RPM performance was studied via dynamic tests performed during the SCINTILLA test and benchmark campaigns. These measurements took place in the JRC ITRAP+10 facility at Ispra (Varese-Italy). The laboratory is equipped with an experimental

  15. A large, high performance, curved 2D position-sensitive neutron detector

    CERN Document Server

    Fried, J W; Mahler, G J; Makowiecki, D S; Mead, J A; Radeka, V; Schaknowski, N A; Smith, G C; Yu, B

    2002-01-01

    A new position-sensitive neutron detector has been designed and constructed for a protein crystallography station at LANL's pulsed neutron source. This station will be one of the most advanced instruments at a major neutron user facility for protein crystallography, fiber and membrane diffraction. The detector, based on neutron absorption in sup 3 He, has a large sensitive area of 3000 cm sup 2 , angular coverage of 120 deg. , timing resolution of 1 mu s, rate capability in excess of 10 sup 6 s sup - sup 1 , position resolution of about 1.5 mm FWHM, and efficiency >50% for neutrons of interest in the range 1-10 A. Features that are key to these remarkable specifications are the utilization of eight independently operating segments within a single gas volume, fabrication of the detector vessel and internal segments with a radius of curvature of about 70 cm, optimized position readout based on charge division and signal shaping with gated baseline restoration, and engineering design with high-strength aluminum ...

  16. The single-collision thermalization approximation for application to cold neutron moderation problems

    International Nuclear Information System (INIS)

    Ritenour, R.L.

    1989-01-01

    The single collision thermalization (SCT) approximation models the thermalization process by assuming that neutrons attain a thermalized distribution with only a single collision within the moderating material, independent of the neutron's incident energy. The physical intuition on which this approximation is based is that the salient properties of neutron thermalization are accounted for in the first collision, and the effects of subsequent collisions tend to average out statistically. The independence of the neutron incident and outscattering energy leads to variable separability in the scattering kernel and, thus, significant simplification of the neutron thermalization problem. The approximation also addresses detailed balance and neutron conservation concerns. All of the tests performed on the SCT approximation yielded excellent results. The significance of the SCT approximation is that it greatly simplifies thermalization calculations for CNS design. Preliminary investigations with cases involving strong absorbers also indicates that this approximation may have broader applicability, as in the upgrading of the thermalization codes

  17. Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

    International Nuclear Information System (INIS)

    Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

    2015-01-01

    Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

  18. Performance of the coupled thermalhydraulics/neutron kinetics code R/P/C on workstation clusters and multiprocessor systems

    International Nuclear Information System (INIS)

    Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.

    1996-01-01

    The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)

  19. The performance and radiation exposure of some neutron probes in measuring the water content of the topsoil layer

    International Nuclear Information System (INIS)

    Arslan, A.; Razzouk, A.K.; Al-Ain, F.

    1997-01-01

    The use of neutron scattering technique for determining the soil surface water content is not popular due to the radiation escaping from the soil surface and the large errors in measurement. To compare the radiation exposure and the performance of different techniques statistically, 3 sites were selected. Five different neutron probe models and different adaptors were used with the depth probes. Exposure to neutrons and γ radiations, at various distances from the probes, were determined. In situ calibration curves were determined using different models of depth probes with a Solo surface reflector block, CPN surface adaptor, and different numbers of plastic Teflon parallelepiped, as well as surface Troxler 3401-B probes. Depth neutron probe readings increased with increasing number of Teflon plastic blocks deposited on the soil surface. The intercept of the straight line regression analysis of CR (count ratio, surface count over standard count) v. percentage water content on a volume basis decreased with increasing number of blocks deposited on the soil surface at all sites. The determination coefficient values of any depth probe with a Solo surface reflector or a block of 4-8 cm thickness were higher than those of a Troxler 3401-B surface probe or CPN 503 depth probe with its surface adaptor. The least exposure to radiation was with a depth probe with surface reflectors. This study proves the possibility of measuring the moisture content of the soil surface by using a depth neutron probe with a block laid on the surface, without danger of receiving the threshold dose of radiation. (authors)

  20. Design of a neutron interrogation cell based on an electron accelerator and performance assessment on 220 liter nuclear waste mock-up drums

    International Nuclear Information System (INIS)

    Sari, A.; Carrel, F.; Laine, F.; Lyoussi, A.

    2013-01-01

    Radiological characterization of nuclear waste drums is an important task for the nuclear industry. The amount of actinides, such as 235 U or 239 Pu, contained in a package can be determined using non-destructive active methods based on the fission process. One of these techniques, known as neutron interrogation, uses a neutron beam to induce fission reactions on the actinides. Optimization of the neutron flux is an important step towards improving this technique. Electron accelerators enable to achieve higher neutron flux intensities than the ones delivered by deuterium-tritium generators traditionally used on neutron interrogation industrial facilities. In this paper, we design a neutron interrogation cell based on an electron accelerator by MCNPX simulation. We carry out photoneutron interrogation measurements on uranium samples placed at the center of 220 liter nuclear waste drums containing different types of matrices. We quantify impact of the matrix on the prompt neutron signal, on the ratio between the prompt and delayed neutron signals, and on the interrogative neutron half-life time. We also show that characteristics of the conversion target of the electron accelerator enable to improve significantly measurement performances. (authors)

  1. Performance of a MICROMEGAS-based TPC in a high-energy neutron beam

    Science.gov (United States)

    Snyder, L.; Manning, B.; Bowden, N. S.; Bundgaard, J.; Casperson, R. J.; Cebra, D. A.; Classen, T.; Duke, D. L.; Gearhart, J.; Greife, U.; Hagmann, C.; Heffner, M.; Hensle, D.; Higgins, D.; Isenhower, D.; King, J.; Klay, J. L.; Geppert-Kleinrath, V.; Loveland, W.; Magee, J. A.; Mendenhall, M. P.; Sangiorgio, S.; Seilhan, B.; Schmitt, K. T.; Tovesson, F.; Towell, R. S.; Walsh, N.; Watson, S.; Yao, L.; Younes, W.

    2018-02-01

    The MICROMEGAS (MICRO-MEsh GAseous Structure) charge amplification structure has found wide use in many detection applications, especially as a gain stage for the charge readout of Time Projection Chambers (TPCs). Here we report on the behavior of a MICROMEGAS TPC when operated in a high-energy (up to 800 MeV) neutron beam. It is found that neutron-induced reactions can cause discharges in some drift gas mixtures that are stable in the absence of the neutron beam. The discharges result from recoil ions close to the MICROMEGAS that deposit high specific ionization density and have a limited diffusion time. For a binary drift gas, increasing the percentage of the molecular component (quench gas) relative to the noble component and operating at lower pressures generally improves stability.

  2. Performance of silicon pad detectors after mixed irradiations with neutrons and fast charged hadrons

    Energy Technology Data Exchange (ETDEWEB)

    Kramberger, G. [Jozef Stefan Institute, Department of Physics, University of Ljubljana, Jamova 39, SI-1000 Ljubljana (Slovenia)], E-mail: Gregor.Kramberger@ijs.si; Cindro, V.; Dolenc, I.; Mandic, I.; Mikuz, M.; Zavrtanik, M. [Jozef Stefan Institute, Department of Physics, University of Ljubljana, Jamova 39, SI-1000 Ljubljana (Slovenia)

    2009-10-11

    A large set of silicon pad detectors produced on MCz and FZ wafer of p- and n-type was irradiated in two steps, first by fast charged hadrons followed by reactor neutrons. In this way the irradiations resemble the real irradiation fields at LHC. After irradiations controlled annealing started in steps during which the evolution of full depletion voltage, leakage current and charge collection efficiency was monitored. The damage introduced by different irradiation particles was found to be additive. The most striking consequence of that is a decrease of the full depletion voltage for n-type MCz detectors after additional neutron irradiation. This confirms that effective donors introduced by charged hadron irradiation are compensated by acceptors from neutron irradiation.

  3. Performance of silicon pad detectors after mixed irradiations with neutrons and fast charged hadrons

    International Nuclear Information System (INIS)

    Kramberger, G.; Cindro, V.; Dolenc, I.; Mandic, I.; Mikuz, M.; Zavrtanik, M.

    2009-01-01

    A large set of silicon pad detectors produced on MCz and FZ wafer of p- and n-type was irradiated in two steps, first by fast charged hadrons followed by reactor neutrons. In this way the irradiations resemble the real irradiation fields at LHC. After irradiations controlled annealing started in steps during which the evolution of full depletion voltage, leakage current and charge collection efficiency was monitored. The damage introduced by different irradiation particles was found to be additive. The most striking consequence of that is a decrease of the full depletion voltage for n-type MCz detectors after additional neutron irradiation. This confirms that effective donors introduced by charged hadron irradiation are compensated by acceptors from neutron irradiation.

  4. Impact of speciation on CO{sub 2} capture performance using blended absorbent containing ammonia, triethanolamine and 2-amino-2-methyl-1-propanol

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hao-Yang; Jeon, Soo-Bin; Jang, Se-Yong; Oh, Kwang-Joong [Pusan National University, Busan (Korea, Republic of); Lee, Sang-Sup [Chungbuk National University, Cheongju (Korea, Republic of); Kang, Seong-Kuy [BK Environmental Construction, Suwon (Korea, Republic of)

    2014-07-15

    In our previous study, a high CO{sub 2} absorption rate was achieved using a blended absorbent containing AMP, NH{sub 3}, and TEA. The species of the blended absorbent was determined in this study using {sup 13}C nuclear magnetic resonance (NMR) spectroscopy and a modified Kent-Eisenberg model. The carbamate formation constant was also regressed using the model. Bicarbonate and carbonate ions decrease the absorption efficiency and have a positive effect on CO{sub 2} stripping. Carbamate has a negative effect on regeneration; a regeneration temperature of 373 K minimized the energy needed. In conclusion, the prediction equation and NMR analysis provide an easy way of determining carbonate group species and carbamate species concentrations, and this method will be helpful in optimizing CO{sub 2} capture with blended absorbents.

  5. Neutronic performance of high molecular weight coolants for a prismatic VHTR

    International Nuclear Information System (INIS)

    Schriener, T. M.; El-Genk, M. S.

    2008-01-01

    A neutronic model is developed of a prismatic Very High Temperature Reactor (VHTR) to investigate the effects on the excess reactivity and operation cycle length of replacing helium with binary gas mixtures of He-Ne, He-N 2 , or He-Xe as reactor coolants and working fluids in the direct Closed Brayton Cycle (CBC) for energy conversion. Also investigated is the neutron activation of these binary gas mixtures in the VHTR. The motivation for using the heavy binary mixtures is the smaller size and the fewer number of stages of the CBC turbo-machinery. The present analysis uses the Monte Carlo code MCNPX 2.6D at typical operating conditions (500-1000 degrees and 7.12 MPa) in the VHTR. He-Ne (15 g/mol) is the best neutronically, but not thermal-hydraulically, followed by He-N 2 . Although He-Ne has ∼13.6% lower heat transfer coefficient than helium, it insignificantly affects the initial excess reactivity and the operation life cycle and experiences no neutrons activation. On the other hand, He-N 2 has 4.4% higher heat transfer coefficient than helium and experiences insignificant neutron activation in the reactor, but decreases the initial excess reactivity by ∼5.2% and the operation cycle length by 6.7%. He-Xe (15 g/mol) has 8% higher heat transfer coefficient than helium, but decreases the initial excess reactivity by 18.2% and the operational cycle length by 17%. In addition, neutron activation of xenon produces a significant source term, requiring shielding of the CBC loop and could contaminate the turbo-machinery with long-lived radioactive cesium. Thus, He-Xe is not recommended as a reactor coolant, but could be used as working fluid in a CBC loop that is indirectly coupled to helium cooled VHTR. (authors)

  6. TPX/TFTR Neutral Beam energy absorbers

    International Nuclear Information System (INIS)

    Dahlgren, F.; Wright, K.; Kamperschroer, J.; Grisham, L.; Lontai, L.; Peters, C.; VonHalle, A.

    1993-01-01

    The present beam energy absorbing surfaces on the TFTR Neutral Beams such as Ion Dumps, Calorimeters, beam defining apertures, and scrapers, are simple water cooled copper plates which wee designed to absorb (via their thermal inertia) the incident beam power for two seconds with a five minute coal down interval between pulses. These components are not capable of absorbing the anticipated beam power loading for 1000 second TPX pulses and will have to be replaced with an actively cooled design. While several actively cooled energy absorbing designs were considered,, the hypervapotron elements currently being used on the JET beamlines were chosen due to their lower cooling water demands and reliable performance on JET

  7. Transportable type neutron level indicators

    International Nuclear Information System (INIS)

    Khatskevich, M.V.; Kalinin, O.V.; Moskovkin, V.N.; Molchanov, A.V.; Bobkov, A.D.; Rabotnov, Yu.A.

    1979-01-01

    Some peculiarities of designing level neutron converters (LNC) for portable indicators or level neutron relays are considered. The effect of the LNC geometry and other factors on measurement errors has been studied. Calibration results of the LNC with a neutron reflector and without it are presented. It is shown that the problem of level monitoring with the help of portable indicators can be solved practically for any volume, provided two LNC modifications with reflectors are available: the NPU-G modification with horizontal location of a counter for large volumes and the NPU-V with vertical location of a counter for lesser volumes. A possibility of perfecting LNC performances by shielding the counter with thermal neutron absorbers has been studied. The design of the NPU-V modification for the NIUP-2 level indicator is described. It is intended for tubes and cylinders 30-100 mm in diameter. Measurements carried out on different steel and aluminium vessels with a diameter ranging from 300 to 100 mm and a wall thickness of up to 16 mm with the help of the NPU-V and NPU-G modifications proved the efficiency of the LNC to control a variety of products (kerosine, gasoline, oils, acids, alkalis) [ru

  8. Neutron radiography

    International Nuclear Information System (INIS)

    Bayon, G.

    1989-01-01

    Neutronography or neutron radiography, a non-destructive test method which is similar in its principle to conventional X-ray photography, presently occupies a marginal position among non-destructive test methods (NDT) (no source of suitable performance or cost). Neutron radiography associated with the ORPHEE reactor permits industrial testing; it can very quickly meet a cost requirement comparable to that of conventional test methods. In 1988, 2500 parts were tested on this unit [fr

  9. Experiments on the Biological Actions of Neutrons Performed in the Former Soviet Union: A Historical Review

    Science.gov (United States)

    2006-10-01

    Radio- biologia , 1982, vol. 22, no. 1, pp. 76-81. 10. Bogatyrev A.V., Timoshenko S.I., Lavrova G.A., Nikanorova N.G., Kalmykova G.I., Sverdlov A.G., On...Chernichenko V.A., Cytogenetic Effectiveness of the Therapeutic Fast Neutron Beam Emitted by Cyclotron U-120. Radio- biologia , vol. 24, no. 4, pp. 567-569, 1986...Dependence of Recessive Sex-Linked Lethal Mu- tations in Drosophila Spermatogenesis on Dose from Fast Neutrons. Radio- biologia , v. 2, no. 3, pp. 420-421

  10. Hydraulic shock absorbers

    International Nuclear Information System (INIS)

    Thatcher, G.; Davidson, D. F.

    1984-01-01

    A hydraulic shock absorber of the dash pot kind for use with electrically conducting liquid such as sodium, has magnet means for electro magnetically braking a stream of liquid discharged from the cylinder. The shock absorber finds use in a liquid metal cooled nuclear reactor for arresting control rods

  11. Personnel neutron dose assessment upgrade: Volume 2, Field neutron spectrometer for health physics applications

    International Nuclear Information System (INIS)

    Brackenbush, L.W.; Reece, W.D.; Miller, S.D.

    1988-07-01

    Both the (ICRP) and the (NCPR) have recommended an increase in neutron quality factors and the adoption of effective dose equivalent methods. The series of reports entitled Personnel Neutron Dose Assessment Upgrade (PNL-6620) addresses these changes. Volume 1 in this series of reports (Personnel Neutron Dosimetry Assessment) provided guidance on the characteristics, use, and calibration of personnel neutron dosimeters in order to meet the new recommendations. This report, Volume 2: Field Neutron Spectrometer for Health Physics Applications describes the development of a portable field spectrometer which can be set up for use in a few minutes by a single person. The field spectrometer described herein represents a significant advance in improving the accuracy of neutron dose assessment. It permits an immediate analysis of the energy spectral distribution associated with the radiation from which neutron quality factor can be determined. It is now possible to depart from the use of maximum Q by determining and realistically applying a lower Q based on spectral data. The field spectrometer is made up of two modules: a detector module with built-in electronics and an analysis module with a IBM PC/reg sign/-compatible computer to control the data acquisition and analysis of data in the field. The unit is simple enough to allow the operator to perform spectral measurements with minimal training. The instrument is intended for use in steady-state radiation fields with neutrons energies covering the fission spectrum range. The prototype field spectrometer has been field tested in plutonium processing facilities, and has been proven to operate satisfactorily. The prototype field spectrometer uses a 3 He proportional counter to measure the neutron energy spectrum between 50 keV and 5 MeV and a tissue equivalent proportional counter (TEPC) to measure absorbed neutron dose

  12. Neutron therapy

    International Nuclear Information System (INIS)

    Riesler, Rudi

    1995-01-01

    Standard radiotherapy uses Xrays or electrons which have low LET (linear energy transfer); in contrast, particles such as neutrons with high LET have different radiobiological responses. In the late 1960s, clinical trials by Mary Catterall at the Hammersmith Hospital in London indicated that fast neutron radiation had clinical advantages for certain malignant tumours. Following these early clinical trials, several cyclotron facilities were built in the 1980s for fast neutron therapy, for example at the University of Washington, Seattle, and at UCLA. Most of these newer machines use extracted cyclotron proton beams in the range 42 to 66 MeV with beam intensities of 15 to 60 microamps. The proton beams are transported to dedicated therapy rooms, where neutrons are produced from beryllium targets. Second-generation clinical trials showed that accurate neutron beam delivery to the tumour site is more critical than for photon therapy. In order to achieve precise beam geometries, the extracted proton beams have to be transported through a gantry which can rotate around the patient and deliver beams from any angle; also the neutron beam outline (''field shape'') must be adjusted to extremely irregular shapes using a flexible collimation system. A therapy procedure has to be appropriately organized, with physicians, radiotherapists, nurses, medical physicists and other staff in attendance; other specialized equipment, such as CT or MRI scanners and radiation simulators must be made available. Neutron therapy is usually performed only in radiation oncology departments of major medical centres

  13. Patterns of Lethality and Absorbed Dose Distributions in Mice for Monoenergetic Neutrons; Letalite et Distribution de la Dose Absorbee chez la Souris pour des Neutrons Monoenergetiques; Letal'nost' i raspredelenie pogloshchennoj dozy pri obluchenii myshej monoehnergeticheskimi neitronami; Letalidad y Distribucion de las Dosis Absorbidas por el Raton para Neutrones Monoenergeticos

    Energy Technology Data Exchange (ETDEWEB)

    Frigerio, N. A.; Jordan, D. L. [Argonne National Laboratory, Argonne, IL (United States)

    1964-03-15

    The presence of strong C, N and O resonances in the 100 to 1500 keV region has permitted the study of specific neutron-nuclide interactions as reflected in lethality, RBE maxima etc. Sixty-two {mu}A of resolved Van de Graaf protons, 1882 to 2738 keV, yielded monoenergetic neutrons via Li{sup 7}(p, n)Be{sup 7}. Virgin female CF-1 mice were exposed in celluloid capsules to the mono-energetic neutrons at distances of 3.1 to 11.3 cm from the source at laboratory angles of 0 to 1 radian. Mice were exposed bilaterally while simultaneously in motion through either circular or elliptical orbits normal to the axis of the beam. Thus, control of dose distribution within the animal was possible. Absolute flux measurements were made with U{sup 235} fission counters and by absolute counting of Au wires and foils activated within Cd covers. Patterns of dose absorption were measured with micro-ionization chambers and with a specially developed FeSO{sub 4}-NH{sub 4}SCN dosimeter of high sensitivity. Relative dose measurements were made with Hurst proton-recoil gas counters and B{sup 10} , Li{sup 6} and proton-recoil scintillators. Neutron-energy distributions were measured with specially developed B{sup 10}, He{sup 3} and Li{sup 6} gas and solid-state spectrometers. Gamma contributions were measured with Ne/Ar chamber counters. These measurements showed gamma contribution to be less than 0.8%, and thermal-epithermal less than 0.01%, of the total rad dose. Animals were exposed to median midpoint doses ranging from 180 to 1200 rad at neutron energies from 396 to 658 keV {+-} 50 keV to cover the region of N and O resonances. Levels and patterns of lethality proved to be strong functions of neutron energy and equally strong, but independent, functions of dose distribution. Regardless of dose, energy or distribution, however, all animals surviving five days survived at least 144 days, dying then of the usual long-term effects. This suggests that monoenergetic fast neutrons, free of

  14. Estimated neutron dose to embryo and foetus during commercial flight

    International Nuclear Information System (INIS)

    Chen, J.; Lewis, B. J.; Bennett, L. G. I.; Green, A. R.; Tracy, B. L.

    2005-01-01

    A study has been carried out to assess the radiation exposure from cosmic-ray neutrons to the embryo and foetus of pregnant aircrew and air travellers in consideration of the radiation exposure from cosmic-ray neutrons to the embryo and foetus. A Monte Carlo analysis was performed to determine the equivalent dose from neutrons to the brain and body of an embryo at 8 weeks and to the foetus at the 3, 6 and 9 month periods. Neutron fluence-to-absorbed dose conversion coefficients for the foetal brain and for the entire foetal body (isotropic irradiation geometry) have been determined at the four developmental stages. The equivalent dose rate to the foetus during commercial flights has been further evaluated considering the fluence-to-absorbed dose conversion coefficients, a neutron spectrum measured at an altitude of 11.3 km and an ICRP-92 radiation-weighting factor for neutrons. This study indicates that the foetus can exceed the annual dose limit of 1 mSv for the general public after, for example, 15 round trips on commercial trans-Atlantic flights. (authors)

  15. Design and performance of high-pressure PLANET beamline at pulsed neutron source at J-PARC

    Energy Technology Data Exchange (ETDEWEB)

    Hattori, T.; Sano-Furukawa, A. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Quantum Beam Science Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Arima, H. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan); Komatsu, K. [Geochemical Research Center, Graduate School of Science, The University of Tokyo, Tokyo 113-0033 (Japan); Yamada, A. [University of Shiga Prefecture, Shiga 522-8533 (Japan); Inamura, Y.; Nakatani, T. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Seto, Y. [Graduate School of Science, Kobe University, Kobe 657-8501 (Japan); Nagai, T. [Faculty of Science, Hokkaido University, Sapporo 060-0810 (Japan); Utsumi, W. [Quantum Beam Science Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Iitaka, T. [Computational Astrophysics Laboratory, RIKEN, Saitama 351-0198 (Japan); Kagi, H. [Geochemical Research Center, Graduate School of Science, The University of Tokyo, Tokyo 113-0033 (Japan); Katayama, Y. [Quantum Beam Science Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Inoue, T. [Geodynamic Research Center, Ehime University, Matsuyama 790-8577 (Japan); Otomo, T. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Institute of Materials Structure Science, High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 205-001 (Japan); Suzuya, K. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Kamiyama, T. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Institute of Materials Structure Science, High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 205-001 (Japan); Arai, M. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Yagi, T. [Geochemical Research Center, Graduate School of Science, The University of Tokyo, Tokyo 113-0033 (Japan)

    2015-04-21

    PLANET is a time-of-flight (ToF) neutron beamline dedicated to high-pressure and high-temperature experiments. The large six-axis multi-anvil high-pressure press designed for ToF neutron diffraction experiments enables routine data collection at high pressures and high temperatures up to 10 GPa and 2000 K, respectively. To obtain clean data, the beamline is equipped with the incident slits and receiving collimators to eliminate parasitic scattering from the high-pressure cell assembly. The high performance of the diffractometer for the resolution (Δd/d~0.6%) and the accessible d-spacing range (0.2–8.4 Å) together with low-parasitic scattering characteristics enables precise structure determination of crystals and liquids under high pressure and temperature conditions.

  16. Design and performance of high-pressure PLANET beamline at pulsed neutron source at J-PARC

    International Nuclear Information System (INIS)

    Hattori, T.; Sano-Furukawa, A.; Arima, H.; Komatsu, K.; Yamada, A.; Inamura, Y.; Nakatani, T.; Seto, Y.; Nagai, T.; Utsumi, W.; Iitaka, T.; Kagi, H.; Katayama, Y.; Inoue, T.; Otomo, T.; Suzuya, K.; Kamiyama, T.; Arai, M.; Yagi, T.

    2015-01-01

    PLANET is a time-of-flight (ToF) neutron beamline dedicated to high-pressure and high-temperature experiments. The large six-axis multi-anvil high-pressure press designed for ToF neutron diffraction experiments enables routine data collection at high pressures and high temperatures up to 10 GPa and 2000 K, respectively. To obtain clean data, the beamline is equipped with the incident slits and receiving collimators to eliminate parasitic scattering from the high-pressure cell assembly. The high performance of the diffractometer for the resolution (Δd/d~0.6%) and the accessible d-spacing range (0.2–8.4 Å) together with low-parasitic scattering characteristics enables precise structure determination of crystals and liquids under high pressure and temperature conditions

  17. Study of the performance of collision short time approximation for neutron scattering using discrete frequency distribution

    International Nuclear Information System (INIS)

    D'Oliveira, A.B.; Amorim, E.S. do; Galvao, O.B.

    1981-03-01

    Double differential cross sections for thermal neutrons, based on incoherent approximation, using continum distribution as discrete frequency set are theoretically estimated, regarding two models previously done. The FASTT computer program is used in order to obtain a numerical estimation. (L.C.) [pt

  18. Neutron radiography

    International Nuclear Information System (INIS)

    Alaa eldin, M.T.

    2011-01-01

    The digital processing of the neutron radiography images gives the possibility for data quantification. In this case an exact relation between the measured neutron attenuation and the real macroscopic attenuation coefficient for every point of the sample is required. The assumption that the attenuation of the neutron beam through the sample is exponential is valid only in an ideal case where a monochromatic beam, non scattering sample and non background contribution are assumed. In the real case these conditions are not fulfilled and in dependence on the sample material we have more or less deviation from the exponential attenuation law. Because of the high scattering cross-sections of hydrogen (σs=80.26 barn) for thermal neutrons, the problem with the scattered neutrons at quantitative radiography investigations of hydrogenous materials (as PE, Oil, H 2 O, etc) is not trivial. For these strong scattering materials the neutron beam attenuation is no longer exponential and a dependence of the macroscopic attenuation coefficient on the material thickness and on the distance between the sample and the detector appears. When quantitative radiography (2 D) or tomography investigations (3 D) are performed, some image correction procedures for a description of the scattering effect are required. This thesis presents a method that can be used to enhance the neutron radiography image for objects with high scattering materials like hydrogen, carbon and other light materials. This method uses the Monte Carlo code, MCNP5, to simulate the neutron radiography process and get the flux distribution for each pixel of the image and determine the scattered neutrons distribution that causes the image blur and then subtract it from the initial image to improve its quality.

  19. Performance test of Si PIN photodiode line scanner for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Totsuka, Daisuke, E-mail: totsuka@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Nihon Kessho Kogaku Co., Ltd., 810-5 Nobe-cho Tatebayashi, Gunma 374-0047 (Japan); Yanagida, Takayuki [New Industry Creation Hatchery Center (NICHe) 6-6-10 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi 980-8579 (Japan); Fukuda, Kentaro; Kawaguchi, Noriaki [Tokuyama Corp., 3 Shibuya Shibuya-ku, Tokyo 150-8383 (Japan); Fujimoto, Yutaka [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Pejchal, Jan [Institute of Physics AS CR, Cukrovarnicka 10, Prague 6, 162-53 (Czech Republic); Yokota, Yuui [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yoshikawa, Akira [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); New Industry Creation Hatchery Center (NICHe) 6-6-10 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi 980-8579 (Japan)

    2011-12-11

    Thermal neutron imaging using Si PIN photodiode line scanner and Eu-doped LiCaAlF{sub 6} crystal scintillator has been developed. The pixel dimensions of photodiode are 1.18 mm (width) Multiplication-Sign 3.8 mm (length) with 0.4 mm gap and the module has 192 channels in linear array. The emission peaks of Eu-doped LiCaAlF{sub 6} after thermal neutron excitation are placed at 370 and 590 nm, and the corresponding photon sensitivities of photodiode are 0.04 and 0.34 A/W, respectively. Polished scintillator blocks with a size of 1.18 mm (width) Multiplication-Sign 3.8 mm (length) Multiplication-Sign 5.0 mm (thickness) were wrapped by several layers of Teflon tapes as a reflector and optically coupled to the photodiodes by silicone grease. JRR-3 MUSASI beam line emitting 13.5 meV thermal neutrons with the flux of 8 Multiplication-Sign 10{sup 5} n/cm{sup 2} s was used for the imaging test. As a subject for imaging, a Cd plate was moved at the speed of 50 mm/s perpendicular to the thermal neutron beam. Analog integration time was set to be 416.6 {mu}s, then signals were converted by a delta-sigma A/D converter. After the image processing, we successfully obtained moving Cd plate image under thermal neutron irradiation using PIN photodiode line scanner coupled with Eu-doped LiCaAlF{sub 6} scintillator.

  20. Performance test of Si PIN photodiode line scanner for thermal neutron detection

    International Nuclear Information System (INIS)

    Totsuka, Daisuke; Yanagida, Takayuki; Fukuda, Kentaro; Kawaguchi, Noriaki; Fujimoto, Yutaka; Pejchal, Jan; Yokota, Yuui; Yoshikawa, Akira

    2011-01-01

    Thermal neutron imaging using Si PIN photodiode line scanner and Eu-doped LiCaAlF 6 crystal scintillator has been developed. The pixel dimensions of photodiode are 1.18 mm (width)×3.8 mm (length) with 0.4 mm gap and the module has 192 channels in linear array. The emission peaks of Eu-doped LiCaAlF 6 after thermal neutron excitation are placed at 370 and 590 nm, and the corresponding photon sensitivities of photodiode are 0.04 and 0.34 A/W, respectively. Polished scintillator blocks with a size of 1.18 mm (width)×3.8 mm (length)×5.0 mm (thickness) were wrapped by several layers of Teflon tapes as a reflector and optically coupled to the photodiodes by silicone grease. JRR-3 MUSASI beam line emitting 13.5 meV thermal neutrons with the flux of 8×10 5 n/cm 2 s was used for the imaging test. As a subject for imaging, a Cd plate was moved at the speed of 50 mm/s perpendicular to the thermal neutron beam. Analog integration time was set to be 416.6 μs, then signals were converted by a delta-sigma A/D converter. After the image processing, we successfully obtained moving Cd plate image under thermal neutron irradiation using PIN photodiode line scanner coupled with Eu-doped LiCaAlF 6 scintillator.

  1. Performance characteristics of a prompt gamma-ray activation analysis (PGAA) system equipped with a new compact D-D neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Joon; Song, Byung Chul; Im, Hee-Jung [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Dukjin-dong 150-1, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Dukjin-dong 150-1, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: kjy@kaeri.re.kr

    2009-07-21

    A new prompt gamma-ray activation analysis (PGAA) system equipped with a compact deuterium-deuterium (D-D) neutron generator has been developed for fast detection of explosives and chemical warfare agents. The PGAA system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF)-driven ion source. The ionic current of the compact neutron generator was determined as a function of the acceleration voltage at various RF powers. Monoenergetic neutrons (2.45 MeV) with a neutron yield of >1x10{sup 7} n/s were obtained at a deuterium pressure of 8.0 mTorr, an acceleration voltage of 80 kV, and an RF power of 1.1 kW. The performance of the PGAA system was examined by studying the dependence of a prompt gamma-ray count rate on crucial operating parameters.

  2. Neutron beams. Tracks analysis, imaging and medicine

    International Nuclear Information System (INIS)

    Pepy, G.

    2006-01-01

    Thermal neutron beams can supply informations about the arrangement of atoms and molecules and about their movement inside the matter. This article treats of the preparation of thermal neutron beams and of the applications that use their penetration and matter activation properties: 1 - thermal neutrons production; 2 - basic properties of thermal neutrons: neutrons scattering, absorbing materials, activating materials, transparent materials, preparation of a neutron beam; 3 - tracks measurement by activation: activation method, measurement of marine pollution by heavy elements, historical evolution of glass composition; 4 - neutron radiography: neutronography, neutronoscopy: viscosity measurement; 5 - cancer treatment. (J.S.)

  3. Determination of absorbed dose in reactors

    International Nuclear Information System (INIS)

    1971-01-01

    There are many areas in the use and operation of research reactors where the absorbed dose and the neutron fluence are required. These include work on the determination of the radiolytic stability of the coolant and moderator and on the determination of radiation damage in structural materials, and reactor experiments involving radiation chemistry and radiation biology. The requirements range from rough estimates of the total heating due to radiation to precise values specifying the contributions of gamma rays, thermal neutrons and fast neutrons. To meet all these requirements a variety of experimental measurements and calculations as well as a knowledge of reactor radiations and their interactions is necessary. Realizing the complexity and importance of this field, its development at widely separated laboratories and the need to bring the experts in this work together, the IAEA has convened three panel meetings. These were: 'In-pile dosimetry', held in July 1964 (published by the Agency as Technical Reports Series No. 46); 'Neutron fluence measurements', in October 1965; and 'In-pile dosimetry', in November 1966. The recommendations of these three panels led the Agency to form a Working Group on Reactor Radiation Measurements and to commission the writing of this book and a book on Neutron Fluence Measurements. The latter was published in May 1970 (Technical Reports Series No. 107). The material on neutron fluence and absorbed dose measurements is widely scattered in reports and reviews. It was considered that it was time for all relevant information to be evaluated and put together in the form of a practical guide that would be valuable both to experienced workers and beginners in the field

  4. Neutron absorption spectroscopy for identification of light elements in actinides

    Energy Technology Data Exchange (ETDEWEB)

    Hau, I.D. [Lawrence Livermore National Laboratory, Advanced Detector Group, 7000 East Ave., L-270, Livermore, CA 94550 (United States) and Department of Nuclear Engineering, University of California Berkeley, Berkeley, CA 94720 (United States)]. E-mail: hau2@llnl.gov; Niedermayr, T.R. [Lawrence Livermore National Laboratory, Advanced Detector Group, 7000 East Ave., L-270, Livermore, CA 94550 (United States); Drury, O.B. [Lawrence Livermore National Laboratory, Advanced Detector Group, 7000 East Ave., L-270, Livermore, CA 94550 (United States); Burger, A. [Fisk University, 1000 17th Ave. North, Nashville, TN 37208 (United States); Bell, Z. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States); Friedrich, S. [Lawrence Livermore National Laboratory, Advanced Detector Group, 7000 East Ave., L-270, Livermore, CA 94550 (United States)]. E-mail: friedrich1@llnl.gov

    2006-04-15

    We are developing cryogenic high-energy resolution fast-neutron spectrometers using superconducting transition-edge sensors (TES) for nuclear science and non-proliferation applications. Fast neutrons are absorbed in 94% enriched {sup 6}LiF single crystals with volumes of {approx}1 cm{sup 3} in an exothermic {sup 6}Li(n,{alpha}){sup 3}H capture reaction. The neutron energy is measured from the subsequent temperature rise with a Mo/Cu multilayer TES. Fast-neutron spectra from a {sup 252}Cf source show an energy resolution of 55 kev. Here, we discuss the instrument performance, with emphasis on the identification of light elements in actinide matrices.

  5. Advances in absorbed dose measurement standards at the australian radiation laboratory

    International Nuclear Information System (INIS)

    Boas, J.F.; Hargrave, N.J.; Huntley, R.B.; Kotler, L.H.; Webb, D.V.; Wise, K.N.

    1996-01-01

    The applications of ionising radiation in the medical and industrial fields require both an accurate knowledge of the amount of ionising radiation absorbed by the medium in question and the capability of relating this to National and International standards. The most useful measure of the amount of radiation is the absorbed dose which is defined as the energy absorbed per unit mass. For radiotherapy, the reference medium is water, even though the measurement of the absorbed dose to water is not straightforward. Two methods are commonly used to provide calibrations in absorbed dose to water. The first is the calibration of the chamber in terms of exposure in a Cobalt-60 beam, followed by the conversion by a protocol into dose to water in this and higher energy beams. The other route is via the use of a graphite calorimeter as a primary standard device, where the conversion from absorbed dose to graphite to absorbed dose in water is performed either by theoretical means making use of cavity ionisation theory, or by experiment where the graphite calorimeter and secondary standard ionisation chamber are placed at scaled distances from the source of the radiation beam (known as the Dose-Ratio method). Extensive measurements have been made at Cobalt-60 at ARL using both the exposure and absorbed dose to graphite routes. Agreement between the ARL measurements and those based on standards maintained by ANSTO and NPL is within ± 0.3%. Absorbed dose measurements have also been performed at ARL with photon beams of nominal energy 16 and 19 MeV obtained from the ARL linac. The validity of the protocols at high photon energies, the validity of the methods used to convert from absorbed dose in graphite to absorbed dose in water and the validity of the indices used to specify the beams are discussed. Brief mention will also be made of the establishment of a calibration facility for neutron monitors at ARL and of progress in the development of ERP dosimetry

  6. A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout

    International Nuclear Information System (INIS)

    Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee

    2012-01-01

    Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.

  7. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S. [Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, Dhaka (Bangladesh); Berliner, R., E-mail: RB@instrumentationAssociates.com [Instrumentation Associates, Durham, North Carolina (United States); Yelon, W. B., E-mail: yelonwb@hotmail.com [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2016-07-12

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  8. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  9. TOMS Absorbing Aerosol Index

    Data.gov (United States)

    Washington University St Louis — TOMS_AI_G is an aerosol related dataset derived from the Total Ozone Monitoring Satellite (TOMS) Sensor. The TOMS aerosol index arises from absorbing aerosols such...

  10. Performance of a parallel algorithm for solving the neutron diffusion equation on the hypercube

    International Nuclear Information System (INIS)

    Kirk, B.L.; Azmy, Y.Y.

    1989-01-01

    The one-group, steady state neutron diffusion equation in two- dimensional Cartesian geometry is solved using the nodal method technique. By decoupling sets of equations representing the neutron current continuity along the length of rows and columns of computational cells a new iterative algorithm is derived that is more suitable to solving large practical problems. This algorithm is highly parallelizable and is implemented on the Intel iPSC/2 hypercube in three versions which differ essentially in the total size of communicated data. Even though speedup was achieved, the efficiency is very low when many processors are used leading to the conclusion that the hypercube is not as well suited for this algorithm as shared memory machines. 10 refs., 1 fig., 3 tabs

  11. Cased Hole Evaluation Using Pulsed Neutron measurements and remedial actions on non-performing wells

    International Nuclear Information System (INIS)

    Mukerji, P.

    2002-01-01

    Advances in pulsed neutron spectroscopy tools have improved accuracy and precision of measured carbon-oxygen rations. The C/O ratios relate to the volumes of oil and water in the formation. Some of the improvements in accuracy and precision have resulted from better tool characterization in a wider variety of logging environments in the calibration facility and new spectral standards. The ability to combine advanced logging measurements has provided the operator with better diagnosis tools for identifying candidates for possible remedial actions. The successful diagnosis and treatment of water production problems requires the identification of specific influx zones. The information obtained from such logs allows effective treatment of unwanted wellbore fluid entries. This paper will present examples from logs run in the Niger delta. We will show how the application of pulsed neutron logs can optimise subsequent well intervention to reduce water production and/or increase oil production

  12. Contribution to a neutronic calculation scheme for pressurized water reactors

    International Nuclear Information System (INIS)

    Martin Del Campo, C.

    1987-01-01

    This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost

  13. Neutronic performance of decoupled poisoned and unpoisoned composite moderators for high resolution experiments

    International Nuclear Information System (INIS)

    Kai, Tetsuya; Harada, Masahide; Watanabe, Noboru; Teshigawara, Makoto; Sakata, Hideaki; Ikeda, Yujiro

    2001-01-01

    We studied decoupled poisoned and un-poisoned composite moderators consisting of 20 mm thick hydrogen and 30 mm thick light water. The neutron pulses from un-poisoned one were much broader with longer decay times than a simple decoupled hydrogen moderator in 50 mm thickness. It was also found that the poisoned composite moderator provides higher pulse peak intensities relative to the hydrogen moderator (poisoned at 20 mm) below several tens meV with no penalty of pulse width. (author)

  14. Neutronics of pulsed spallation neutron sources

    International Nuclear Information System (INIS)

    Watanabe, Noboru

    2003-01-01

    Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, poisoning, etc are discussed, aiming at a high performance pulsed spallation source

  15. Neutron radiography for nondestructive testing

    International Nuclear Information System (INIS)

    John, J.

    1979-01-01

    Neutron radiography is similar to X-ray inspection in that both depend upon use of radiation that penetrates some materials and is absorbed by others to provide a contrast image of conditions not readily available for visual inspection. X-rays are absorbed by dense materials, such as metals, whereas neutrons readily penetrate metals, but are absorbed by materials containing hydrogen. The neutron radiography has been successfully applied to a number of inspection situations. These include the inspection of explosives, advanced composites, adhesively bonded structures and a number of aircraft engine components. With the availability of Californium-252, it has become feasible to construct mobile neutron radiography systems suitable for field use. Such systems have been used for in-situ inspection of flight line aircraft, particularly to locate and measure hidden corrosion

  16. Performance of the prototype LANL solid deuterium ultra-cold neutron source

    CERN Document Server

    Hill, R E; Bowles, T J; Greene, G L; Hogan, G; Lamoreaux, S; Marek, L; Mortenson, R; Morris, C L; Saunders, A; Seestrom, S J; Teasdale, W A; Hoedl, S; Liu, C Y; Smith, D A; Young, A; Filippone, B W; Hua, J; Ito, T; Pasyuk, E A; Geltenbort, P; García, A; Fujikawa, B; Baessler, S; Serebrov, A

    2000-01-01

    A prototype of a solid deuterium (SD sub 2) source of Ultra-Cold Neutrons (UCN) is currently being tested at LANSCE. The source is contained within an assembly consisting of a 4 K polyethylene moderator surrounded by a 77 K beryllium flux trap in which is embedded a spallation target. Time-of-flight measurements have been made of the cold neutron spectrum emerging directly from the flux trap assembly. A comparison is presented of these measurements with results of Monte Carlo (LAHET/MCNP) calculations of the cold neutron fluxes produced in the prototype assembly by a beam of 800 MeV protons incident on the tungsten target. A UCN detector was coupled to the assembly through a guide system with a critical velocity of 8 m/s ( sup 5 sup 8 Ni). The rates and time-of-flight data from this detector are compared with calculated values. Measurements of UCN production as a function of SD sub 2 volume (thickness) are compared with predicted values. The dependence of UCN production on SD sub 2 temperature and proton beam...

  17. Enhancing the performances of a resonance detector spectrometer for deep inelastic neutron scattering measurements

    International Nuclear Information System (INIS)

    Filabozzi, A.; Pace, E.; Pietropaolo, A.

    2012-01-01

    The possibility is explored to sum up neutron Compton profiles at different scattering angles in deep inelastic neutron scattering measurements within the Resonance Detector (RD) configuration to enhance the statistics for a more reliable extraction of the momentum distribution of the constituents in the target. The RD configuration allows to select the energy of the scattered neutrons up to several tens of electron Volt, thus accessing energy and wave vector transfers well above 1 eV and 30 Å −1 , respectively. In the high-q/ω regime, the final state effects could be considered as negligible, as shown in a series of simulations using a Monte Carlo method with different inverse geometry instrument setups. The simulations show that it could be possible to conceive an instrument set up where the RD configuration allows the proper summation of several spectra at different scattering angles, providing a good separation of the proton recoil signal from that of the heavier atoms, thus avoiding the cell subtraction by fitting procedure.

  18. Performance improvement of an active vibration absorber subsystem for an aircraft model using a bees algorithm based on multi-objective intelligent optimization

    Science.gov (United States)

    Zarchi, Milad; Attaran, Behrooz

    2017-11-01

    This study develops a mathematical model to investigate the behaviour of adaptable shock absorber dynamics for the six-degree-of-freedom aircraft model in the taxiing phase. The purpose of this research is to design a proportional-integral-derivative technique for control of an active vibration absorber system using a hydraulic nonlinear actuator based on the bees algorithm. This optimization algorithm is inspired by the natural intelligent foraging behaviour of honey bees. The neighbourhood search strategy is used to find better solutions around the previous one. The parameters of the controller are adjusted by minimizing the aircraft's acceleration and impact force as the multi-objective function. The major advantages of this algorithm over other optimization algorithms are its simplicity, flexibility and robustness. The results of the numerical simulation indicate that the active suspension increases the comfort of the ride for passengers and the fatigue life of the structure. This is achieved by decreasing the impact force, displacement and acceleration significantly.

  19. Installation and development of neutron radiography in the nuclear reactor (IEAR-1) of the Instituto de Energia Atomica, Brazil

    International Nuclear Information System (INIS)

    Fuga, R.

    1979-01-01

    Investigations on the field of Neutron Radiography have been performed at the IEAR-1, swimming pool reactor utilizing a collimated neutron beam and the so-called photographic transfer method as a mean of detection. The test object (sample) is placed between the neutron source (reactor core) and the gold foil. The acitivity of its different points is the inverse measure of the neutrons absorbed in the test sample at the corresponding points. The activity distribution on the gold foil is determined again by exposing it to an X-ray film. A multichannel type collimator consisting of an assemblage of stainless steel tubes inside an aluminium mantle (tube) was used as a direction beam selector. Improvements have been introduced in respect to the reduction of angular divergence and neutron scattering. To improve further the quality of the radiographs another collimator type has been developed using boric acid as a neutron absorber and moderator. Flux measurements by means of gold foil activation at reactor positions of interest were necessary to eliminate errors originating of different neutron flux values. The dependence of film darkening upon the neutron flux and other factors have been discussed. Finally neutron-and gama-radiographs of the same objects were evaluated in comparison. (author) [pt

  20. Very slow neutrons

    International Nuclear Information System (INIS)

    Frank, A.

    1983-01-01

    The history is briefly presented of the research so far of very slow neutrons and their basic properties are explained. The methods are described of obtaining very slow neutrons and the problems of their preservation are discussed. The existence of very slow neutrons makes it possible to perform experiments which may deepen the knowledge of the fundamental properties of neutrons. Their wavelength approximates that of visible radiation. The possibilities and use are discussed of neutron optical systems (neutron microscope) which could be an effective instrument for the study of the detailed arrangement, especially of organic substances. (B.S.)

  1. The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

    Energy Technology Data Exchange (ETDEWEB)

    Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.

    2014-10-01

    The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.

  2. Liquid absorber experiments in ZED-2

    International Nuclear Information System (INIS)

    McDonnell, F.N.

    1975-07-01

    A set of liquid absorber experiments was performed in ZED-2 to provide data with which to test the adequacy of calculational methods for zone controller and refuelling studies associated with advanced reactor concepts. The absorber consisted of a full length aluminum tube, containing either i)H 2 O, ii)H 2 O + boron (2.5 mg/ml) or iii)H 2 O + boron (8.0 mg/ml). The tube was suspended vertically at interstitial or in-channel locations. A U-tube absorber was also simulated using two absorber tubes with appropriate spacers. Experiments were carried out at two different square lattice pitches, 22.86 and 27.94 cm. Measurements were made of the reactivity effects of the absorbers and, in some cases, of the detailed flux distribution near the perturbation. The results from one calculational method, the source-sink approach, were compared with the data from selected experiments. (author)

  3. Dispersion-Type Absorbing Materials for the Control Organs of Thermal Reactors; Absorbants du Type a Dispersion pour les Organes de Commande des Reacteurs a Neutrons Thermiques; Pogloshchayushchie materialy dispersionnogo tipa dlya organov regulirovaniya teplovykh reaktorov; Absorbentes de Tipo Dispersion para los Organos de Mando de los Reactores Termicos

    Energy Technology Data Exchange (ETDEWEB)

    Nosov, V. I.; Ponomarjov-Stepnoj, H. H.; Portnoj, K. I.; Savel' ev, E. G.

    1964-06-15

    The paper gives the results of a study of the physical characteristics of NIMONIC-type absorbing alloys with oxides of rare-earth elements dispersed in them (gadolinium, samarium, europium etc. ). The paper discusses changes in absorbing capacity in relation to the composition of the material, describes the mechanical and thermophysical properties of the absorbing materials as a function of the concentration of absorber introduced into the alloy and, finally, gives the results of a study of the effect of radiation on the properties of the materials. It is shown that absorbing alloys with oxides of rare-earth elements dispersed in the metallic matrix possess considerable absorbing capacity for relatively small amounts of absorber in the alloy (5 to 10%). When oxides of rare-earth elements are added, NIMONIC-type alloys have relatively high resistance and thermophysical characteristics (o{sub B}, E, {lambda}) at high temperatures for absorber concentrations up to about 10%. Dispersion materials of this type have satisfactory radiation stability in a radiation field of about 3 x 10{sup 20}n/cm{sup 2} at high temperature. (author) [French] Les auteurs exposent les resultats de recherches sur les caracteristiques physiques des alliages absorbants du type nimonik, contenant des terres rares dispersees dans leur masse (gadolinium, samarium, europium, etc.). Ils examinent les variations de la capacite d'absorption selon la composition du materiau; on donne des indications sur les caracteristiques mecaniques et thermophysiques des absorbants en fonction de la concentration de Tabsorbeur incorpore dans l 'alliage ainsi que les resultats d 'une etude relative a l 'influence de l'irradiation sur ces caracteristiques. Ils montrent que les alliages absorbants contenant des oxydes de terres rares disperses dans une matrice metallique ont une capacite d'absorption importante pour une teneur de l'alliage relativement faible en'matieres absorbantes (environ 5 a 10%). Les alliages du

  4. Fundamental of neutron radiography and the present of neutron radiography in Japan

    International Nuclear Information System (INIS)

    Sekita, Junichiro

    1988-01-01

    Neutron radiography refers to the application of transmitted neutrons to analysis. In general, thermal neutron is used for neutron radiography. Thermal neutron is easily absorbed by light atoms, including hydrogen, boron and lithium, while it is not easily absorbed by such heavy atoms as tungsten, lead and uranium, permitting detection of impurities in heavy metals. Other neutrons than thermal neutron can also be applied. Cold neutron is produced from fast neutron using a moderator to reduce its energy down to below that of thermal neutron. Cold neutron is usefull for analysis of thick material. Epithermal neutron can induce resonance characteristic of each substance. With a relatively small reaction area, fast neutron permits observation of thick samples. Being electrically neutral, neutrons are difficult to detect by direct means. Thus a substance that releases charged particles is put in the path of neutrons for indirect measurement. X-ray film combined with converter screen for conversion of neutrons to charge particles is placed behind the sample. Photographing is carried out by a procedure similar to X-ray photography. Major institues and laboratories in Japan provided with neutron radiography facilities are listed. (Nogami, K.)

  5. Design and development of radiation absorber for sighting beam line

    International Nuclear Information System (INIS)

    Sridhar, R.; Shukla, S.K.

    2005-01-01

    During the commissioning of Indus-2 , it is necessary to view the synchrotron radiation that will be emanating from the dipole exit ports. The 10 0 beam line from dipole 11 was earmarked for sighting beam line. The synchrotron radiation power density would be around 340 watts on the photon absorber inside the radiation absorber module, at the specified beam power of Indus-2. The beam striking on this photon absorber produces x-rays and Bremsstrahlung radiation. These are to be stopped and absorbed by radiation absorber. The photon absorber and the radiation absorber are integrated in a single vacuum chamber and actuated by a pneumatic cylinder connected using a bellow. Radiation absorber was needed to isolate the diagnostic components and to protect them from radiation a well as heat when they were not in use. The paper describes the design, calculation and development of the dynamic photon cum radiation absorber. The ultimate vacuum performance is also described. (author)

  6. Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

  7. Evaluation of the dark signal performance of different SiPM-technologies under irradiation with cold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Durini, Daniel, E-mail: d.durini@fz-juelich.de [Central Institute of Engineering, Electronics and Analytics ZEA-2 – Electronic Systems, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Degenhardt, Carsten; Rongen, Heinz [Central Institute of Engineering, Electronics and Analytics ZEA-2 – Electronic Systems, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Feoktystov, Artem [Jülich Centre for Neutron Science (JCNS) at Heinz Maier-Leibnitz Zentrum (MLZ), Forschungszentrum Jülich GmbH, Lichtenbergstr. 1, D-85748 Garching (Germany); Schlösser, Mario; Palomino-Razo, Alejandro [Central Institute of Engineering, Electronics and Analytics ZEA-2 – Electronic Systems, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Frielinghaus, Henrich [Jülich Centre for Neutron Science (JCNS) at Heinz Maier-Leibnitz Zentrum (MLZ), Forschungszentrum Jülich GmbH, Lichtenbergstr. 1, D-85748 Garching (Germany); Waasen, Stefan van [Central Institute of Engineering, Electronics and Analytics ZEA-2 – Electronic Systems, Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany)

    2016-11-01

    In this paper we report the results of the assessment of changes in the dark signal delivered by three silicon photomultiplier (SiPM) detector arrays, fabricated by three different manufacturers, when irradiated with cold neutrons (wavelength λ{sub n}=5 Å or neutron energy of E{sub n}=3.27 meV) up to a neutron dose of 6×10{sup 12} n/cm{sup 2}. The dark signals as well as the breakdown voltages (V{sub br}) of the SiPM detectors were monitored during the irradiation. The system was characterized at room temperature. The analog SiPM detectors, with and without a 1 mm thick Cerium doped {sup 6}Li-glass scintillator material located in front of them, were operated using a bias voltage recommended by the respective manufacturer for a proper detector performance. I{sub out}-V{sub bias} measurements, used to determine the breakdown voltage of the devices, were repeated every 30 s during the first hour and every 300 s during the rest of the irradiation time. The digital SiPM detectors were held at the advised bias voltage between the respective breakdown voltage and dark count mappings repeated every 4 min. The measurements were performed on the KWS-1 instrument of the Heinz Maier-Leibnitz Zentrum (MLZ) in Garching, Germany. The two analog and one digital SiPM detector modules under investigation were respectively fabricated by SensL (Ireland), Hamamatsu Photonics (Japan), and Philips Digital Photon Counting (Germany).

  8. Developmental testing of partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Allen, G.C.; Rack, H.J.; Joseph, B.J.; Dupree, S.A.

    1980-01-01

    Results of the phase one tests have demonstrated that the neutron-shielding concept described in this paper is a viable design option for spent fuel shipping casks. The tests have shown that the Boro-silicone 236 shield is superior to the other shield materials considered. Repeated TGA, aging and fire tests demonstrated the reliability of the data. A second phase of the test program is now being pursued where the Boro-silicone 236 is injected into all-steel slab sections, and cured in place. 5 tables

  9. Performance of neutron activation analysis in the evaluation of bismuth iodide purification methodology

    International Nuclear Information System (INIS)

    Armelin, Maria Jose A.; Ferraz, Caue de Mello; Hamada, Margarida M.

    2015-01-01

    Bismuth tri-iodide (BrI 3 ) is an attractive material for using as a semiconductor. In this paper, BiI 3 crystals have been grown by the vertical Bridgman technique using commercially available powder. The impurities were evaluated by instrumental neutron activation analysis (INAA). The results show that INAA is an analytical method appropriate for monitoring the impurities of: Ag, As, Br, Cr, K, Mo, Na and Sb in the various stages of the BiI 3 purification methodology. (author)

  10. Dosimetry and biological effects of fast neutrons

    International Nuclear Information System (INIS)

    Zoetelief, J.

    1981-01-01

    This thesis contains studies on two types of cellular damage: cell reproductive death and chromosome aberrations induced by irradiation with X rays, gamma rays and fast neutrons of different energies. A prerequisite for the performance of radiobiological experiments is the determination of the absorbed dose with a sufficient degree of accuracy and precision. Basic concepts of energy deposition by ionizing radiation and practical aspects of neutron dosimetry for biomedical purposes are discussed. Information on the relative neutron sensitivity of GM counters and on the effective point of measurement of ionization chambers for dosimetry of neutron and photon beams under free-in-air conditions and inside phantoms which are used to simulate the biological objects is presented. Different methods for neutron dosimetry are compared and the experimental techniques used for the investigations of cell reproductive death and chromosome aberrations induced by ionizing radiation of different qualities are presented. Dose-effect relations for induction cell inactivation and chromsome aberrations in three cultured cell lines for different radiation qualities are presented. (Auth.)

  11. Radiation portal monitor with {sup 10}B+ZnS(Ag) neutron detector performance for the detection of special nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Guzman G, K. A.; Gallego, E.; Lorente, A.; Ibanez F, S. [Universidad Politecnica de Madrid, Departamento de Ingenieria Energetica, ETSI Industriales, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Gonzalez, J. A. [Universidad Politecnica de Madrid, Laboratorio de Ingenieria Nuclear, ETSI Caminos, Canales y Puertos, C. Prof. Aranguren 3, 28040 Madrid (Spain); Mendez, R., E-mail: ingkarenguzman@gmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Laboratorio de Patrones Neutronicos, Av. Complutense 40, 28040 Madrid (Spain)

    2016-10-15

    In homeland security, neutron detection is used to prevent the smuggling of special nuclear materials. Thermal neutrons are normally detected with {sup 3}He proportional counters, in the radiation portal monitors, Rpms, however due to the {sup 3}He shortage new procedures are being studied. In this work Monte Carlo methods, using the MCNP6 code, have been used to study the neutron detection features of a {sup 10}B+ZnS(Ag) under real conditions inside of a Rpm. The performance for neutron detection was carried out for {sup 252}Cf, {sup 238}U and {sup 239}Pu under different conditions. In order to mimic an actual situation occurring at border areas, a sample of SNM sited inside a vehicle was simulated and the Rpm with {sup 10}B+ZnS(Ag) response was calculated. At 200 cm the {sup 10}B+ZnS(Ag) on Rpm response is close to 2.5 cps-ng {sup 252}Cf, when the {sup 252}Cf neutron source is shielded with 0.5 cm-thick lead and 2.5 cm-thick polyethylene fulfilling the ANSI recommendations. Three different geometries of neutron detectors of {sup 10}B+ZnS(Ag) in a neutron detection system in Rpm were modeled. Therefore, the {sup 10}B+ZnS(Ag) detectors are an innovative and viable replacement for the {sup 3}He detectors in the Rpm. (Author)

  12. Radiation portal monitor with "1"0B+ZnS(Ag) neutron detector performance for the detection of special nuclear materials

    International Nuclear Information System (INIS)

    Guzman G, K. A.; Gallego, E.; Lorente, A.; Ibanez F, S.; Vega C, H. R.; Gonzalez, J. A.; Mendez, R.

    2016-10-01

    In homeland security, neutron detection is used to prevent the smuggling of special nuclear materials. Thermal neutrons are normally detected with "3He proportional counters, in the radiation portal monitors, Rpms, however due to the "3He shortage new procedures are being studied. In this work Monte Carlo methods, using the MCNP6 code, have been used to study the neutron detection features of a "1"0B+ZnS(Ag) under real conditions inside of a Rpm. The performance for neutron detection was carried out for "2"5"2Cf, "2"3"8U and "2"3"9Pu under different conditions. In order to mimic an actual situation occurring at border areas, a sample of SNM sited inside a vehicle was simulated and the Rpm with "1"0B+ZnS(Ag) response was calculated. At 200 cm the "1"0B+ZnS(Ag) on Rpm response is close to 2.5 cps-ng "2"5"2Cf, when the "2"5"2Cf neutron source is shielded with 0.5 cm-thick lead and 2.5 cm-thick polyethylene fulfilling the ANSI recommendations. Three different geometries of neutron detectors of "1"0B+ZnS(Ag) in a neutron detection system in Rpm were modeled. Therefore, the "1"0B+ZnS(Ag) detectors are an innovative and viable replacement for the "3He detectors in the Rpm. (Author)

  13. Digital Alloy Absorber for Photodetectors

    Science.gov (United States)

    Hill, Cory J. (Inventor); Ting, David Z. (Inventor); Gunapala, Sarath D. (Inventor)

    2016-01-01

    In order to increase the spectral response range and improve the mobility of the photo-generated carriers (e.g. in an nBn photodetector), a digital alloy absorber may be employed by embedding one (or fraction thereof) to several monolayers of a semiconductor material (insert layers) periodically into a different host semiconductor material of the absorber layer. The semiconductor material of the insert layer and the host semiconductor materials may have lattice constants that are substantially mismatched. For example, this may performed by periodically embedding monolayers of InSb into an InAsSb host as the absorption region to extend the cutoff wavelength of InAsSb photodetectors, such as InAsSb based nBn devices. The described technique allows for simultaneous control of alloy composition and net strain, which are both key parameters for the photodetector operation.

  14. A neutronics study for improving the safety and