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Sample records for neutronics design studies

  1. Conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at SINQ

    International Nuclear Information System (INIS)

    Zanini, L.; Baluc, N.; Simone, A. De; Eichler, R.; Joray, S.; Manfrin, E.; Pouchon, M.; Rabaioli, S.; Schumann, D.; Welte, J.; Zhernosekov, K.

    2011-12-01

    This comprehensive, illustrated report by the Paul Scherrer Institute PSI in Switzerland documents the proposals concerning the conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at the PSI's Swiss Spallation Neutron Source SINQ facility. The need for fast neutron irradiation is discussed and the possibility of using SINQ as a fast neutron irradiation facility is considered. The production of isotopes, tracers and medical isotopes is discussed, as are fission and fusion reactor technologies. The characteristics of the neutron spectrum in SINQ are discussed. The neutronic and radioprotection calculations for an irradiation station at SINQ are looked at in detail and extensive examples of work done and results obtained are presented and discussed. Radioprotection issues are also looked at. Further contributions in the report cover the hot/cold irradiation station in the SINQ target. An appendix provides detailed drawings of the facility's pneumatic delivery system

  2. Linac design study for an intense neutron-source driver

    International Nuclear Information System (INIS)

    Lynch, M.T.; Browman, A.; DeHaven, R.; Jameson, R.; Jason, A.; Neuschaefer, G.; Tallerico, P.; Regan, A.

    1993-01-01

    The 1-MW spallation-neutron source under design study at Los Alamos is driven by a linac-compressor-ring scheme that utilizes a large portion of the existing Los Alamos Meson Physics Facility (LAMPF) linac, as well as the facility infrastructure. The project is referred to as the National Center for Neutron Research (NCNR). A second phase of the proposal will upgrade the driver power to 5 MW. A description of the 1-MW scheme is given in this paper. In addition, the upgrade path to the substantial increase of beam power required for the 5 MW scenario is discussed

  3. Neutronic design studies for an unattended, low power reactor

    International Nuclear Information System (INIS)

    Palmer, R.G.; Durkee, J.W. Jr.

    1986-01-01

    The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported

  4. 5 MW pulsed spallation neutron source, Preconceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This report describes a self-consistent base line design for a 5 MW Pulsed Spallation Neutron Source (PSNS). It is intended to establish feasibility of design and as a basis for further expanded and detailed studies. It may also serve as a basis for establishing project cost (30% accuracy) in order to intercompare competing designs for a PSNS not only on the basis of technical feasibility and technical merit but also on the basis of projected total cost. The accelerator design considered here is based on the objective of a pulsed neutron source obtained by means of a pulsed proton beam with average beam power of 5 MW, in {approx} 1 {mu}sec pulses, operating at a repetition rate of 60 Hz. Two target stations are incorporated in the basic facility: one for operation at 10 Hz for long-wavelength instruments, and one operating at 50 Hz for instruments utilizing thermal neutrons. The design approach for the proton accelerator is to use a low energy linear accelerator (at 0.6 GeV), operating at 60 Hz, in tandem with two fast cycling booster synchrotrons (at 3.6 GeV), operating at 30 Hz. It is assumed here that considerations of cost and overall system reliability may favor the present design approach over the alternative approach pursued elsewhere, whereby use is made of a high energy linear accelerator in conjunction with a dc accumulation ring. With the knowledge that this alternative design is under active development, it was deliberately decided to favor here the low energy linac-fast cycling booster approach. Clearly, the present design, as developed here, must be carried to the full conceptual design stage in order to facilitate a meaningful technology and cost comparison with alternative designs.

  5. 5 MW pulsed spallation neutron source, Preconceptual design study

    International Nuclear Information System (INIS)

    1994-06-01

    This report describes a self-consistent base line design for a 5 MW Pulsed Spallation Neutron Source (PSNS). It is intended to establish feasibility of design and as a basis for further expanded and detailed studies. It may also serve as a basis for establishing project cost (30% accuracy) in order to intercompare competing designs for a PSNS not only on the basis of technical feasibility and technical merit but also on the basis of projected total cost. The accelerator design considered here is based on the objective of a pulsed neutron source obtained by means of a pulsed proton beam with average beam power of 5 MW, in ∼ 1 μsec pulses, operating at a repetition rate of 60 Hz. Two target stations are incorporated in the basic facility: one for operation at 10 Hz for long-wavelength instruments, and one operating at 50 Hz for instruments utilizing thermal neutrons. The design approach for the proton accelerator is to use a low energy linear accelerator (at 0.6 GeV), operating at 60 Hz, in tandem with two fast cycling booster synchrotrons (at 3.6 GeV), operating at 30 Hz. It is assumed here that considerations of cost and overall system reliability may favor the present design approach over the alternative approach pursued elsewhere, whereby use is made of a high energy linear accelerator in conjunction with a dc accumulation ring. With the knowledge that this alternative design is under active development, it was deliberately decided to favor here the low energy linac-fast cycling booster approach. Clearly, the present design, as developed here, must be carried to the full conceptual design stage in order to facilitate a meaningful technology and cost comparison with alternative designs

  6. A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

    A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

  7. A Time of Flight Fast Neutron Imaging System Design Study

    Science.gov (United States)

    Canion, Bonnie; Glenn, Andrew; Sheets, Steven; Wurtz, Ron; Nakae, Les; Hausladen, Paul; McConchie, Seth; Blackston, Matthew; Fabris, Lorenzo; Newby, Jason

    2017-09-01

    LLNL and ORNL are designing an active/passive fast neutron imaging system that is flexible to non-ideal detector positioning. It is often not possible to move an inspection object in fieldable imager applications such as safeguards, arms control treaty verification, and emergency response. Particularly, we are interested in scenarios which inspectors do not have access to all sides of an inspection object, due to interfering objects or walls. This paper will present the results of a simulation-based design parameter study, that will determine the optimum system design parameters for a fieldable system to perform time-of-flight based imaging analysis. The imaging analysis is based on the use of an associated particle imaging deuterium-tritium (API DT) neutron generator to get the time-of-flight of radiation induced within an inspection object. This design study will investigate the optimum design parameters for such a system (e.g. detector size, ideal placement, etc.), as well as the upper and lower feasible design parameters that the system can expect to provide results within a reasonable amount of time (e.g. minimum/maximum detector efficiency, detector standoff, etc.). Ideally the final prototype from this project will be capable of using full-access techniques, such as transmission imaging, when the measurement circumstances allow, but with the additional capability of producing results at reduced accessibility.

  8. Design study of a medical proton linac for neutron therapy

    International Nuclear Information System (INIS)

    Machida, S.; Raparia, D.

    1988-01-01

    This paper describes a design study which establishes the physical parameters of the low energy beam transport, radiofrequency quadrupole, and linac, using computer programs available at Fermilab. Beam dynamics studies verify that the desired beam parameters can be achieved. The machine described here meets the aforementioned requirements and can be built using existing technology. Also discussed are other technically feasible options which could be attractive to clinicians, though they would complicate the design of the machine and increase construction costs. One of these options would allow the machine to deliver 2.3 MeV protons to produce epithermal neutrons for treating brain tumors. A second option would provide 15 MeV protons for isotope production. 21 refs., 33 figs

  9. Passive neutron design study for 200-L waste drums

    International Nuclear Information System (INIS)

    Menlove, H.O.; Beddingfield, D.B.; Pickrell, M.M.

    1997-09-01

    We have developed a passive neutron counter for the measurement of plutonium in 200-L drums of scrap and waste. The counter incorporates high efficiency for the multiplicity counting in addition to the traditional coincidence counting. The 252 Cf add-a-source feature is used to provide an accurate assay over a wide range of waste matrix materials. The room background neutron rate is reduced by using 30 cm of external polyethylene shielding and the cosmic-ray background is reduced by statistical filtering techniques. Monte Carlo Code calculations were used to determine the optimum detector design, including the gas pressure, size, number, and placement of the 3 He tubes in the moderator. Various moderators, including polyethylene, plastics, teflon, and graphite, were evaluated to obtain the maximum efficiency and minimum detectable mass of plutonium

  10. Thermal design study of a liquid hydrogen-cooled cold-neutron source

    International Nuclear Information System (INIS)

    Quach, D.; Aldredge, R.C.; Liu, H.B.; Richards, W.J.

    2007-01-01

    The use of both liquid hydrogen as a moderator and polycrystalline beryllium as a filter to enhance cold neutron flux at the UC Davis McClellan Nuclear Radiation Center has been studied. Although, more work is needed before an actual cold neutron source can be designed and built, the purpose of this preliminary study is to investigate the effects of liquid hydrogen and the thickness of a beryllium filter on the cold neutron flux generated. Liquid hydrogen is kept at 20 K, while the temperature of beryllium is assumed to be 77 K in this study. Results from Monte Carlo simulations show that adding a liquid hydrogen vessel around the beam tube can increase cold neutron flux by more than an order of magnitude. As the thickness of the liquid hydrogen layer increases up to about half an inch, the flux of cold neutrons also increases. Increasing the layer thickness to more than half an inch gives no significant enhancement of cold neutron flux. Although, the simulations show that the cold neutron flux is almost independent of the thickness of beryllium at 77 K, the fraction of cold neutrons does drop along the beam tube. This may be due to the fact that the beam tube is not shielded for neutrons coming directly from the reactor core. Further design studies are necessary for to achieve complete filtering of undesired neutrons. A simple comparison analysis based on heat transfer due to neutron scattering and gamma-ray heating shows that the beryllium filter has a larger rate of change of temperature and its temperature is higher. As a result heat will be transferred from beryllium to liquid hydrogen, so that keeping liquid hydrogen at the desired temperature will be the most important step in the cooling process

  11. Spallation neutron source moderator design

    International Nuclear Information System (INIS)

    Charlton, L.A.; Barnes, J.M.; Gabriel, T.A.; Johnson, J.O.

    1998-01-01

    This paper describes various aspects of the spallation neutron source (SNS) moderator design. Included are the effects of varying the moderator location, interaction effects between moderators, and the impact on neutron output when various reflector materials are used. Also included is a study of the neutron output from composite moderators, where it is found that a combination of liquid H 2 O and liquid H 2 can produce a spectrum very similar to liquid methane (L-CH 4 ). (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  13. Design of intense neutron source for fusion material study and the role of universities

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1993-01-01

    Need and requirement for the intense neutron source for fusion materials study have been discussed for many years. Recently, international climate has been becoming gradually maturing to consider this problem more seriously because of the recognition of crucial importance of solving materials problems for fusion energy development. The present symposium was designed to discuss the problems associated with the intense neutron source for material irradiation studies which will have a potential for the National Institute for Fusion Science to become one of the important future research areas. The symposium comprises five sessions; first, the role of materials research in fusion development strategies was discussed followed by a brief summary of current IFMIF (International Fusion Materials Irradiation Facility) activity. Despite the pressing need for intense fusion neutron source, currently available neutron sources are reactor or accelerator based sources of which FFTF and LASREF were discussed. Then, various concepts of intense neutron source candidates were presented including ESNIT, which are currently under design by JAERI. In the fourth session, discussions were made on the study of materials with the intense neutron source from the viewpoint of materials scientists and engineers as the user of the facility. This is followed by discussions on the role of universities from the two stand points, namely, fusion irradiation studies and fusion materials development. Finally summary discussions were made by the participants, indicating important role fundamental studies in universities for the full utilization of irradiation data and the need of pure 14 MeV neutron source for fundamental studies together with the intense surrogate neutron sources. (author)

  14. Study of spectral response of a neutron filter. Design of a method to adjust spectra

    International Nuclear Information System (INIS)

    Colomb-Dolci, F.

    1999-02-01

    The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

  15. Design study of a medical reactor for Boron Neutron Capture Therapy

    International Nuclear Information System (INIS)

    Sasaki, M.; Hirota, J.; Tamao, S.; Kanda, K.; Mishima, Y.

    1992-01-01

    A new design study of a medical reactor for Boron Neutron Capture Therapy (BNCT) has been carried out. The reactor is to be used exclusively for the treatment of malignant melanoma and other cancers as well as for the further biomedical research. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of triangular lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam hole are to be provided to deliver thermal and epithermal neutrons. N-γ coupling Sn transport calculations indicate that the patient treatment period will be about 30 minutes with minimal fast neutron and gamma contaminants. (author)

  16. Design studies for a high-resolution, transportable neutron radiography/radioscopy system

    International Nuclear Information System (INIS)

    Gillespie, G.H.; Micklich, B.J.; McMichael, G.E.

    1996-01-01

    A preliminary design has been developed for a high-resolution, transportable neutron radiology system (TNRS) concept. The primary system requirement is taken to be a thermal neutron flux of 10[sup 6] n/(cm[sup 2]-sec) with a L/D ratio of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed,and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, low energy beam transport (LEBT), RFQ, high energy beam transport (HEBT), target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. The use of trade studies for optimizing the TNRS concept are also described

  17. Structural design study of tritium breeding blanket with a lead layer as a neutron multiplier

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kitamura, Kazunori; Minato, Akio; Sakamoto, Hiroki; Yamamoto, Takashi

    1980-12-01

    Thermal and structural design study of a tritium breeding blanket with a lead layer for a International Tokamak Reactor (INTOR) is carried out. Tube in shell type blanket with a lead layer is found to be promising. The volume fraction of structural material in the lead layer can be small enough to keep the neutron multiplication effect of lead. Reasonable value of shell effect is attainable due to lead layer in the front part of the blanket. (author)

  18. Design and construction of a fast (neutron) spectrometer for magnetic confinement fusion studies

    International Nuclear Information System (INIS)

    Lagattu, A.

    1980-09-01

    Neutron emission in conjunction with the 'Tokamak' devices studied in order to carry out controlled fusion is done at present on a short time scale (a few hundred milliseconds). Obtaining spectrometric information on this emission made it necessary to design sufficiently fast equipment to adapt to this situation. The basis of the equipment selected is conventional: NE213 liquid scintillator in conjunction with a fast photomultiplier. A commercialized fast neutron gamma discriminator (approximately 210 5 S -1 ) was used. The data acquisition and processing system was specially made for this occasion. This assembly, so designed and constructed, was tested on a particle accelerator and on the TFR Tokamak. In the best conditions a spectrum of 60 milliseconds was produced. Hence several spectra can be recorded during an experiment. The rejection of gamma radiations by the discriminator exceeds 99% in such conditions [fr

  19. Shielding design study for the JAERI/KEK spallation neutron source

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Teshigawara, Makoto; Konno, Chikara; Ikeda, Yujiro; Watanabe, Noboru

    2001-01-01

    Shielding design for the JAERI/KEK spallation neutron source was studied. Bulk shielding characteristics and optimization of a beam shutter were investigated by using Monte Carlo calculation code NMTC/JAM and MCNP with LA-150 neutron cross-section library. The following remarks were derived. (1) Neutron dose outside of the concrete shield at 6.6 m from the center is ∼10 μSv/hr regardless of angles with respect to the proton beam axis. The neutron dose can be reduced more than a factor of 30 by adding natural boron of 5 wt% in the concrete. (2) When a beam shutter position just outside the void vessel and the shutter length of 2 m are assumed, a shutter made of copper (1.7 m) with polyethylene (0.3 m) is the optimum in terms of shielding performance as well as cost merit. A shutter made of tungsten is not so effective. (3) Further studies are needed for optimization of beam shutter position. (author)

  20. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  1. Design study of superconducting sextupole magnet using HTS coated conductor for neutron-focusing device

    International Nuclear Information System (INIS)

    Tosaka, T.; Koyanagi, K.; Ono, M.; Kuriyama, T.; Watanabe, I.; Tsuchiya, K.; Suzuki, J.; Adachi, T.; Shimizu, H.M.

    2006-01-01

    We performed a design study of sextupole magnet using high temperature superconducting (HTS) wires. The sextupole magnet is used as a focusing lens for neutron-focusing devices. A neutron-focusing device is desired to have a large aperture and a high magnetic field gradient of G, where G = 2B/r 2 , B is the magnetic field and r is a distance from the sextupole magnet axis. Superconducting magnets offer promising prospects to meet the demands of a neutron-focusing device. Recently NbTi coils of low temperature superconducting (LTS) have been developed for a sextupole magnet with a 46.8 mm aperture. The maximum magnetic field gradient G of this magnet is 9480 T/m 2 at 4.2 K and 12,800 T/m 2 at 1.8 K. On the other hand, rapid progress on second generation HTS wire has been made in increasing the performance of critical current and in demonstrating a long length. The second generation HTS wire is referred to as coated conductor. It consists of tape-shaped base upon which a thin coating of superconductor, usually YBCO, is deposited or grown. This paper describes a design study of sextupole magnet using coated conductors

  2. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    Powers, J.

    2008-01-01

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials (1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF 4 or ThF 4 or some combination thereof. Future systems could look at using PuF 3 or PuF 4 as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies

  3. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory

  4. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor

    International Nuclear Information System (INIS)

    Vendryes, G.; Zaleski, C.P.

    1964-01-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios (∼1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 Δk/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [fr

  5. Preliminary neutronics design studies for a 400 MWt STAR-LM

    International Nuclear Information System (INIS)

    Aliberti, G.; Yang, W. S.; Stillman, J. A.; Hill, R. N.

    2004-01-01

    Neutronics design studies for a 400 MWt high temperature fast reactor are being performed, utilizing lead coolant, transuranic (TRU) nitride fuel, and HT-9 structural material. Under the main design constraints of long fuel lifetime, natural convection heat transport, semi-autonomous control, and small unit size, parametric studies were performed to maximize the discharge burnup and minimize the burnup reactivity swing. Based on the results of these parametric studies, two point designs were developed for a single-batch once-through fuel cycle; one is a 15 full power year cycle design with core volume of 9.5 cubic meters, and the other is a 12 full power year cycle design with core volume of 7.4 cubic meters. For these two point designs, fuel cycle analyses and reactivity feedback coefficients calculations were performed. The 9.5 cubic meter design achieved an average discharge burnup of 83 MWd/kg with a maximum reactivity change over the lifetime of 0.6%. The peak fast fluence was well within the fast fluence limit of HT9, and both average and peak power densities were well below the estimated limit for natural circulation. The performances of the 7.4 cubic meter design were slightly inferior to this design. To enhance the passive safety characteristics, however, further design improvements need to be made to reduce the coolant density coefficient and to increase the radial expansion coefficient. (authors)

  6. Increasing the neutron flux study for the TRR-II core design

    International Nuclear Information System (INIS)

    Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

    1999-01-01

    The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

  7. Neutronic moderator design for the Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Charlton, L.A.; Barnes, J.M.; Johnson, J.O.; Gabriel, T.A.

    1998-01-01

    Neutronics analyses are now in progress to support the initial selection of moderator design parameters for the Spallation Neutron Source (SNS). The results of the initial optimization studies involving moderator poison plate location, moderator position, and premoderator performance for the target system are presented in this paper. Also presented is an initial study of the use of a composite moderator to produce a liquid methane like spectrum

  8. A Neutronic Feasibility Study of an OPR-1000 Core Design with Boron-bearing Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Park, Sang Yoon; Lee, Chung Chan; Yang, Yong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Westinghouse plants, boron is mainly used as a form of the integral fuel burnable absorber (IFBA) with a thin coating of zirconium diboride (ZrB{sub 2}) or wet annular burnable absorber (WABA) with a hollow Al{sub 2}O{sub 3}+B{sub 4}C pellet. In OPR-1000, on the other hand, gadolinia is currently employed as a form of an admixture which consists of Gd{sub 2}O{sub 3} of 6∼8 w/o and UO{sub 2} of natural uranium. Recently, boron-bearing UO{sub 2} fuel (BBF) with the high density of greater than 94%TD has been developed by using a low temperature sintering technique. In this paper, the feasibility of replacing conventional gadolinia-bearing UO{sub 2} fuel (GBF) in OPR-1000 with newly developed boron-bearing fuel is evaluated. Neutronic feasibility study to utilize the BBF in OPR-1000 core has been performed. The results show that the OPR-1000 core design with the BBF is feasible and promising in neutronic aspects. Therefore, the use of the BBF in OPR-1000 can reduce the dependency on the rare material such as gadolinium. However, the burnout of the {sup 10}B isotope results in helium gas, so fuel performance related study with respect to helium generation is needed.

  9. Neutron shielding studies on an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    Merk, Bruno; Konheiser, Jörg

    2014-01-01

    Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

  10. Structural design study of a proton beam window for a 1-MW spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Teraoku, Takuji; Terada, Atsuhiko; Maekawa, Fujio; Meigo, Shin-ichiro; Kaminaga, Masanori; Ishikura, Syuichi; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    A 1-MW spallation neutron source aiming at materials and life science researches will be constructed under the JAERI-KEK High-intensity Proton Accelerator Project (J-PARC). A proton beam passes through a proton beam window, and be injected into a target of the neutron source. The proton beam window functions as a boundary wall between a high vacuum area in the proton beam line and a helium atmosphere at about atmospheric pressure in a helium vessel which contains the target and moderators. The proton beam window is cooled by light water because high heat-density is generated in the window material by interactions with the proton beam. Then, uniformity of the water flow is requested at the window to suppress a hot-spot that causes excessive thermal stress and cooling water boiling. Also, the window has to be strong enough in its structure for inner stress due to water pressure and thermal stress due to heat generation. In this report, we propose two types of proton beam windows; one flat-type that is easy to manufacture, and the other, curved-type that has high stress resistivity. As a part of design study for the windows, evaluation of strength of structure and thermal hydraulic analysis were conducted. As a result, it was found that sufficient heat removal was assured with uniform water flow at the window, and stress caused by internal water pressure and thermal stress could be maintained below allowable stress values. Accordingly, it was confirmed that the proton beam window designs were feasible. (author)

  11. Multipurpose intense 14 MeV neutron source at Bratislava: Design study

    International Nuclear Information System (INIS)

    Pivarc, J.; Hlavac, S.; Kral, J.; Oblozinsky, P.; Ribansky, I.; Turzo, I.

    1980-05-01

    The present state of design of the multipurpose intense 14 MeV neutron source based on a D + ion beam and a metal tritide target is reported. It is essentially a 300 keV electrostatic air insulated accelerator capable to accelerate a deuterium ion beam up to 10 mA. With such a beam and a beam spot of 1 cm 2 , a neutron yield typically 10 12 n/s and a useful target lifetime of around 10 h are expected. Various users requirements are met by means of three beam lines: an intense, low current dc and a low current fast pulsed. The key components of the intense source section are the rotating target and the ion source. The rotating target is proposed, with respect of the heat dissipation and the removal of 3 kW/cm 2 , in continuous operation. A rotation speed up to 1100 rpm is considered. The ion source should deliver about 0.5 kW of extracted D + ion beam power. A duoplasmatron source with an electrostatic beam focusing system has been selected. Low current sections of the neutron source may operate with a high frequency ion source as well. The dc section for maximum yields around 10 10 n/s is designed with special regard to beam monitoring. The fast pulsed section should produce up to 1 ns compressible pulsed D + ion beam on a target spot with 5 MHz repetition rate. The report includes information about other components of the neutron source as a high voltage power supply, a vacuum system, beam transport, a diagnostic and control system and basic information about neutron source cells and radiation protection. (author)

  12. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G.; Zaleski, C.P. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les

  13. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G; Zaleski, C P [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les projets de reacteurs futurs

  14. Neutron response study

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Fix, J.J.; Thorson, M.R.; Nichols, L.L.

    1981-01-01

    Neutron response of the albedo type dosimeter is strongly dependent on the energy of the incident neutrons as well as the moderating material on the backside of the dosimeter. This study characterizes the response of the Hanford dosimeter for a variety of neutron energies for both a water and Rando phantom (a simulated human body consisting of an actual human skeleton with plastic for body muscles and certain organs). The Hanford dosimeter response to neutrons of different energies is typical of albedo type dosimeters. An approximate two orders of magnitude difference in response is observed between neutron energies of 100 keV and 10 MeV. Methods were described to compensate for the difference in dosimeter response between a laboratory neutron spectrum and the different spectra encountered at various facilities in the field. Generally, substantial field support is necessary for accurate neutron dosimetry

  15. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    Science.gov (United States)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  16. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  17. Engineering design of a neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, Daniel M.; Campos, Tarcísio P.R. de, E-mail: dmcoelho.eng@gmail.com, E-mail: tprcampos@pq.cnpq.br [Universidade Federal de Minas Gerais (NRI/UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear. Nucleo de Radiações Ionizantes

    2017-07-01

    This paper presents an engineering design of a neutron generator (NG). In order to analyze and choose the materials and the appropriate geometry, previous studies of NRI Group (Nucleus for Ionizing Radiation at UFMG - NRI/UFMG) were considered and a model was developed for the simulation of these systems. The efficiency of a neutron generator is measured by the neutron flux. Among the modeling and simulation methods, was employed open software sources for the transmuting cell, aiming to evaluate resonant cavity and for complementary physical analysis. In addition, the titanium target was compared designed based in other studies of NRI Group. Deuterium plasma with a density close to 10{sup 10} particles/cm³, was proposed with a frequency of 0.898 GHz and an approximate wavelength of 110 μm, using a radio frequency antenna up to 2.45 GHz. This compact system includes a hydrogen-isotopes fusor, moderator, reflector and shield. Neutron reflection minimized the neutron escape, increasing the final flux. A insulation material is required to enclose the device. As a conclusion, the investigated nuclear and electromagnetic features of NG have demonstrated that such generator shall have a notable potential for radioisotope generation applied to medical diagnosis. The designs presented will be used to build a 3D model in the NRI laboratory and then a prototype with the selected materials. (author)

  18. Neutronics studies for the design of the European DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

    2016-11-01

    Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

  19. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    International Nuclear Information System (INIS)

    Liang, C.; Ji, W.

    2013-01-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  20. Neutronic design of an ADS

    International Nuclear Information System (INIS)

    Cintas, A; Lopasso, E.M; Marquez Damian, J.I

    2009-01-01

    We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations and different control rods types in critical or subcritical configurations with an external source. To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic 252 C f source, and a D-D pulsed neutron source. Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified. [es

  1. Design of a linear neutron source

    International Nuclear Information System (INIS)

    Buzarbaruah, N.; Dutta, N.J.; Bhardwaz, J.K.; Mohanty, S.R.

    2015-01-01

    Highlights: • This paper reports the design of a linear neutron source based on inertial electrostatic confinement fusion scheme. • The voltage and current that is to be applied to the grid is computed theoretically. • Neutron production rate is theoretically estimated and found to be of the order of 10 7 –10 8 neutrons/s. • Electric potential distribution and ion trajectories are studied using SIMION code. • Optimized condition for the inner grid transparency has been found out. - Abstract: In this paper, we present the design of a linear neutron source based on the concept of inertial electrostatic confinement fusion. The source mainly comprises of a concentric coaxial cylindrical grid assembly housed inside a double walled cylindrical vacuum chamber, a gas injection system, a high voltage feedthrough and a high voltage negative polarity power supply. The inner grid will be kept at a high negative potential with respect to the outer grid that will be grounded. The effect of grid transparency on electric potential distribution and ion trajectories has been studied using SIMION. A diffuse deuterium plasma will be initially created by making filament discharge and subsequently, on application of high negative voltage to the inner grid, deuterons will be accelerated towards the axis of the device. These deuterons will oscillate in the negative potential and consequently fuse in between the grids to produce neutrons. This source is expected to produce 10 7 –10 8 neutrons/s. The proposed linear neutron source will be operated both in the continuous and pulse modes and it will be utilized for a few near term applications namely fusion reactor material studies and explosive detection

  2. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  3. Design study of a time-of-flight neutron spectrometer for JT-60U

    International Nuclear Information System (INIS)

    Elevant, T.; Hoek, M.; Nishitani, Takeo.

    1993-06-01

    A time-of-flight neutron spectrometer is proposed for measurements of neutron energy spectra from deuterium-deuterium reactions in JT-60U tokamak plasmas. The sensitivity of the instrument is 2 · 10 -2 cm 2 , energy resolution is 4.5 % (FWHM) and maximum useful count-rate is 6 kHz. Analysis of neutron energy spectra will provide information on central ion temperatures larger than ∼ 4 keV with an accuracy of ± 10 %, and neutron source fraction from reactions between thermal ions with an accuracy of ± 15 %. The minimum time required for data acquisition is 0.1 s. (author)

  4. Design of neutron diagnostic for MTX

    International Nuclear Information System (INIS)

    Ogawa, Toshihide; Oasa, Kazumi; Hoshino, Katsumichi; Odajima, Kazuo; Maeda, Hikosuke

    1990-07-01

    A neutron diagnostic system was designed for the Microwave Tokamak Experiment being carried out at the lawrence Livermore National Laboratory. High speed measurements are important to this experiment. Plastic scintillator is used for this fast response detection of neutron. Proportional counters and fission counters are used for the total neutron emission rate measurements. (author)

  5. Experiment Design and Analysis Guide - Neutronics & Physics

    Energy Technology Data Exchange (ETDEWEB)

    Misti A Lillo

    2014-06-01

    The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

  6. Conceptual design of HANARO cold neutron source

    International Nuclear Information System (INIS)

    Lee, Chang Hee; Sim, Cheul Muu; Park, K. N.; Choi, Y. H.

    2002-07-01

    The purpose of the cold source is to increase the available neutron flux delivered to instruments at wavelength 4 ∼ 12 A. The major engineering targets of this CNS facility is established for a reach out of very high gain factors in consideration with the cold neutron flux, moderator, circulation loop, heat load, a simplicity of the maintenance of the facility, safety in the operation of the facility against the hydrogen explosion and a layout of a minimum physical interference with the present facilities. The cold source project has been divided into 5 phases: (1) pre-conceptual (2) conceptual design (3) Testing (4) detailed design and procurement (5) installation and operation. Although there is sometime overlap between the phases, in general, they are sequential. The pre-conceptual design and concept design of KCNS has been performed on elaborations of PNPI Russia and review by Technicatome, Air Liquid, CILAS France. In the design of cold neutron source, the characteristics of cold moderators have been studied to obtain the maximum gain of cold neutron, and the analysis for radiation heat, design of hydrogen system, vacuum system and helium system have been performed. The possibility for materialization of the concept in the proposed conceptual design has been reviewed in view of securing safety and installing at HANARO. Above all, the thermosiphon system to remove heat by circulation of sub-cooled two phase hydrogen has been selected so that the whole device could be installed in the reactor pool with the reduced volume. In order to secure safety, hydrogen safety has been considered on protection to prevent from hydrogen-oxygen reaction at explosion of hydrogen-oxygen e in the containment. A lay out of the installation, a maintenance and quality assurance program and a localization are included in this report. Requirements of user, regulatory, safety, operation, maintenance should be considered to be revised for detailed design, testing, installation

  7. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  8. Neutronics design study on a minor actinide burner for transmuting spent fuel

    International Nuclear Information System (INIS)

    Choi, Hang Bok

    1998-08-01

    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors. The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the doppler coefficient, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics. (author). 34 refs., 22 tabs., 14 figs

  9. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  10. The intense neutron generator study

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1966-07-01

    The study has confirmed that a beam of 65 mA of protons at 1000 MeV, striking a molten lead-bismuth target surrounded by heavy water moderator, would give the desired flux of 10{sup 16} thermal neutrons per cm{sup 2} per second to provide intense beams of neutrons and also to produce radioisotopes. The proton beam passing through a thin auxiliary target would also produce beams of mesons. The design and construction of the ion source, injector, accelerator, target and auxiliaries present challenging technical problems. Moreover, continued development for improved life and economy promises to be rewarding. The high neutron intensity is sought for research in solid and liquid state physics and also for nuclear physics. Participation by universities and industry, both in development and use, is expected to be extensive. (author)

  11. The intense neutron generator study

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1966-01-01

    The study has confirmed that a beam of 65 mA of protons at 1000 MeV, striking a molten lead-bismuth target surrounded by heavy water moderator, would give the desired flux of 10 16 thermal neutrons per cm 2 per second to provide intense beams of neutrons and also to produce radioisotopes. The proton beam passing through a thin auxiliary target would also produce beams of mesons. The design and construction of the ion source, injector, accelerator, target and auxiliaries present challenging technical problems. Moreover, continued development for improved life and economy promises to be rewarding. The high neutron intensity is sought for research in solid and liquid state physics and also for nuclear physics. Participation by universities and industry, both in development and use, is expected to be extensive. (author)

  12. Design and fabrication of 4π Clover Detector Array Assembly for gamma-spectroscopy studies using thermal neutrons

    International Nuclear Information System (INIS)

    Kumar, Manish; Kamble, S.R.; Chaudhari, A.T.; Sabharwal, T.P.; Pathak, Kavindra; Prasad, N.K.; Kinage, L.A.; Biswas, D.C.; Bhagwat, P.V.

    2017-01-01

    Nuclear spectroscopy has been studied earlier from the measurement of prompt gamma rays produced in reactions with thermal neutrons from CIRUS reactor. For studying the prompt γ-spectroscopy using thermal neutrons from Dhruva Reactor, BARC, the development of a dedicated beam line (R-3001) is in progress. In this beam line a detector assembly consisting of Clover Ge detectors will be used. This experimental setup will be utilized to investigate nuclear structure using prompt (n,γ) reactions and also to study the spectroscopy of neutron-rich fission-fragment nuclei

  13. Moderator design studies for a new neutron reference source based on the D–T fusion reaction

    International Nuclear Information System (INIS)

    Mozhayev, Andrey V.; Piper, Roman K.; Rathbone, Bruce A.; McDonald, Joseph C.

    2016-01-01

    The radioactive isotope Californium-252 ( 252 Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D 2 O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252 Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuteriumtritium (D–T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252 Cf. To be viable, the 14 MeV D–T output spectrum must be significantly moderated to approximate common workplace environments. This paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2–5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered. - Highlights: • D–T generator neutron calibration field replacement for D 2 O-moderated 252 Cf. • Determination of representative nuclear power plant workplace neutron spectrum. • Simulations to assess moderating materials to soften 14

  14. Design of neutron detectors utilising luminescent glass

    International Nuclear Information System (INIS)

    Spowart, A.R.

    1983-01-01

    Impetus for the development of new neutron detector designs has derived from the worldwide commissioning of neutron spallation sources. The design concepts, and principal methods of utilisation of these major installations, have been recently reviewed. Their principal feature of interest is their broadband neutron emission allowing neutron investigations of all types of structure in materials from biological molecules to steels. Conventional neutron detectors are gas-filled devices, based on BF/sub 3/ or /sup 3/He gas. Their major advantage is their intrinsically low background count. Their principal disadvantage is their slow response time (10-100 μs), high cost and relative lack of flexibility in design to cope with large areas or complex geometry detection. They are, however, long established and the research facilities around the world have a heavy investment in the interpretative hardware for gas detectors

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

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

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

  16. Simulation study of a hydrostat design for detecting underground leakage of water supply using neutron backscattering

    International Nuclear Information System (INIS)

    Kurosawa, Tadahiro; Nakamura, Takashi; Suzuki, Takashi; Okano, Yasuhiro; Chisaka, Haruo

    1998-01-01

    We have embarked upon the development of a new detection method for underground water leakage using a neutron backscattering system. We have estimated the performance capabilities of such a system using Monte Carlo simulation. It is indicated that a leak which results in 40% water content in the surrounding soils could be detected at depths of up to 40 cm from the surface to the center of the source of leakage. This new detection system could be useful as a hydrostat of underground water supply in noisy areas such as Tokyo, in place of presently-used hydrostats which are based on detection of changes in sound

  17. Monte Carlo design study of a moderated {sup 252}Cf source for in vivo neutron activation analysis of aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, D.G.; Natto, S.S.A.; Evans, C.J. [Swansea In Vivo Analysis and Cancer Research Group, Department of Physics, University of Wales, Swansea (United Kingdom); Ryde, S.J.S. [Swansea In Vivo Analysis and Cancer Research Group, Department of Medical Physics and Clinical Engineering, Singleton Hospital, Swansea (United Kingdom)

    1997-04-01

    The Monte Carlo computer code MCNP has been used to design a moderated 2{sup 52}Cf neutron source for in vivo neutron activation analysis of aluminium (Al) in the bones of the hand. The clinical motivation is the need to monitor l body burden in subjects with renal dysfunction, at risk of Al toxicity. The design involves the source positioned on the central axis at one end of a cylindrical deuterium oxide moderator. The moderator is surrounded by a graphite reflector, with the hand inserted at the end of the moderator opposing the source. For a 1 mg {sup 252}Cf source, 15 cm long x 20 cm radius moderator and 20 cm thick reflector, the estimated minimum detection limit is .5 mg Al for a 20 min irradiation, with an equivalent dose of 16.5 mSv to the hand. Increasing the moderator length and/or introducing a fast neutron filter (for example silicon) further reduces interference from fast-neutron-induced reactions on phosphorus in bone, at the expense of decreased fluence of the thermal neutrons which activate Al. Increased source strengths may be necessary to compensate for this decreased thermal fluence, or allow measurements to be made within an acceptable time limit for the comfort of the patient. (author)

  18. Advanced Neutron Source enrichment study

    International Nuclear Information System (INIS)

    Bari, R.A.; Ludewig, H.; Weeks, J.R.

    1996-01-01

    A study has been performed of the impact on performance of using low-enriched uranium (20% 235 U) or medium-enriched uranium (35% 235 U) as an alternative fuel for the Advanced Neutron Source, which was initially designed to use uranium enriched to 93% 235 U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology

  19. Structural design study of a proton beam window for a 1-MW spallation neutron source

    CERN Document Server

    Teraoku, T; Ishikura, S; Kaminaga, M; Maekawa, F; Meigo, S I; Terada, A

    2003-01-01

    A 1-MW spallation neutron source aiming at materials and life science researches will be constructed under the JAERI-KEK High-intensity Proton Accelerator Project (J-PARC). A proton beam passes through a proton beam window, and be injected into a target of the neutron source. The proton beam window functions as a boundary wall between a high vacuum area in the proton beam line and a helium atmosphere at about atmospheric pressure in a helium vessel which contains the target and moderators. The proton beam window is cooled by light water because high heat-density is generated in the window material by interactions with the proton beam. Then, uniformity of the water flow is requested at the window to suppress a hot-spot that causes excessive thermal stress and cooling water boiling. Also, the window has to be strong enough in its structure for inner stress due to water pressure and thermal stress due to heat generation. In this report, we propose two types of proton beam windows; one flat-type that is easy to m...

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

    CERN Document Server

    Matsumoto, T

    2003-01-01

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

  1. Design study and heat transfer analysis of a neutron converter target for medical radioisotope production

    International Nuclear Information System (INIS)

    Masoud Behzad; Sang-In Bak; Seung-Woo Hong; Jong-Seo Chai; Yacine Kadi; Claudio Tenreiro; University of Talca, Talca

    2014-01-01

    A worldwide challenge in the near future will be to find a way of producing radioisotopes in sufficient quantity without relying on research reactors. The motivation for this innovative work on targets lies in the accelerator-based production of radioisotopes using a neutron converter target as in the transmutation by adiabatic resonance crossing concept. Thermal analysis of a multi-channel helium cooled device is performed with the computational fluid dynamics code CFX. Different boundary conditions are taken into account in the simulation process and many important parameters such as maximum allowable solid target temperature as well as uniform inlet velocity and outlet pressure changes in the channels are investigated. The results confirm that the cooling configuration works well; hence such a solid target could be operated safely and may be considered for a prototype target. (author)

  2. Study on a new design of Tehran Research Reactor for radionuclide production based on fast neutrons using MCNPX code.

    Science.gov (United States)

    Zandi, Nadia; Afarideh, Hossein; Aboudzadeh, Mohammad Reza; Rajabifar, Saeed

    2018-02-01

    The aim of this work is to increase the magnitude of the fast neutron flux inside the flux trap where radionuclides are produced. For this purpose, three new designs of the flux trap are proposed and the obtained fast and thermal neutron fluxes compared with each other. The first and second proposed designs were a sealed cube contained air and D 2 O, respectively. The results of calculated production yield all indicated the superiority of the latter by a factor of 55% in comparison to the first proposed design. The third proposed design was based on changing the surrounding of the sealed cube by locating two fuel plates near that. In this case, the production yield increased up to 70%. Copyright © 2017. Published by Elsevier Ltd.

  3. A new design of neutron survey instrument

    International Nuclear Information System (INIS)

    Tanner, R.J.; Eakins, J.S.; Hager, L.G.

    2010-01-01

    A novel design of neutron survey instrument has been developed. The moderator has been modified via the use of 'neutron guides', which help thermal neutrons reach the central proportional counter. This innovation has allowed the variations in the energy dependence of ambient dose equivalent response to be reduced compared to prior single-detector designs, whilst maintaining a relatively light moderator and simple construction. In particular, the design has a relatively small over-response to neutrons with energies around 5 keV, when compared to prior designs. The final optimized design has been verified using MCNP5 calculations to ensure that the response is relatively independent of the energy and direction of the incident neutron. This has required the ends of the guides to be structured so that unidirectional and isotropic neutron fields have closely matched responses, as is necessary in the workplace. The reading of the instrument in workplace fields is calculated via folding and the suitability of the design for use in the workplace discussed.

  4. Design and Fabrication of Titanium Target for Portable Neutron Generator

    International Nuclear Information System (INIS)

    Lee, Cheol Ho; Oh, Byunghoon; Chang, Daesik; Jang, Dohyun; In Sang Yeol; Park, Jaewon; Hong, Kwangpyo

    2014-01-01

    For the neutron generator to produce a neutron flux of the above order, a target that produces fast neutrons in the generator plays an important role, and the target is used and applied to develop the generator due to its simplicity and inexpensive. Making suitable targets for neutron production, especially mono-energy neutrons, has always been of interest. These targets have been used for neutron production reaction studies, calibration of detectors, and neutron therapy. Different studies have been carried out on deuterium and tritium for making solid targets to produce mono-energy neutron from D-D and D-T reactions. A lot of investigations have been carried out on solid target properties such as lifetime, thermal stability, neutron yield, and energy. Vaporized zirconium and titanium layers on a high thermal conductivity substrate (Cu, Mo, Ag) have been used as deuterium and tritium absorbing metals. The density of titanium is smaller than zirconium and the range of charged particles in the titanium targets is more than that in zirconium targets. Thus, titanium targets have more neutron yield than zirconium targets in a low energy beam and titanium is usually used to make a target. The titanium target was designed and simulated to determine the suitable thickness of the target. As a result of the simulation, the target was fabricated to generate fast neutrons by the reaction. The thickness of the target was measured using a profiler. The thickness of the two targets is 2.108 and 2.190 μm. The target will be applied to produce neutrons in a neutron generator

  5. Neutronic Design Calculations on Moderators for the Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Murphy, D.B.

    1999-01-01

    The Spallation Neutron Source (SNS) to be built at the Oak Ridge National Laboratory will provide an intense source of neutrons for a large variety of experiments. It consists of a high-energy (1-GeV) and high-power (∼1-MW) proton accelerator, an accumulator ring, together with a target station and an experimental area. In the target itself, the proton beam will produce neutrons via the spallation process and these will be converted to low-energy ( 2 O moderators. Extensive engineering design work has been conducted on the moderator vessels. For our studies we have produced realistic neutronic representations of these moderators. We report on neutronic studies conducted on these representations of the moderators using Monte Carlo simulation techniques

  6. Optimization of a neutron detector design using adjoint transport simulation

    International Nuclear Information System (INIS)

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-01-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  7. Advanced Neutron Source: The designer's perspective

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1990-01-01

    The Advanced Neutron Source (ANS) is a research facility based on a 350 MW beam reactor, to be brought into service at the Oak Ridge National Laboratory at the end of the century. The primary objective is to provide high-flux neutron beams and guides, with cold, thermal, hot, and ultra-cold neutrons, for research in many fields of science. Secondary objectives include isotopes production, materials irradiation and activation analysis. The design of the ANS is strongly influenced by the historical development of research and power reactor concepts, and of the regulatory infrastructure of the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC). Current trends in reactor safety also impact the climate for the design of such a reactor

  8. Advanced Neutron Sources: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS

  9. Thermal neutron source study

    International Nuclear Information System (INIS)

    Holden, T.M.

    1983-05-01

    The value of intense neutron beams for condensed matter research is discussed with emphasis on the complementary nature of steady state and pulsed neutron sources. A large body of information on neutron sources, both existing and planned, is then summarized under four major headings: fission reactors, electron accelerators with heavy metal targets, pulsed spallation sources and 'steady state' spallation sources. Although the cost of a spallation source is expected to exceed that of a fission reactor of the same flux by a factor of two, there are significant advantages for a spallation device such as the proposed Electronuclear Materials Test Facility (EMTF)

  10. Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

    Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

  11. Neutron scattering studies in the actinide region

    International Nuclear Information System (INIS)

    Kegel, G.H.R.; Egan, J.J.

    1993-09-01

    This report discusses the following topics: Prompt fission neutron energy spectra for 235 U and 239 Pu; Two-parameter measurement of nuclear lifetimes; ''Black'' neutron detector; Data reduction techniques for neutron scattering experiments; Inelastic neutron scattering studies in 197 Au; Elastic and inelastic scattering studies in 239 Pu; and neutron induced defects in silicon dioxide MOS structures

  12. Neutron response study using PADC

    International Nuclear Information System (INIS)

    El-Badry, B.A; Hegazy, T.M; Morsy, A.A.; Zaki, M.F.

    2007-01-01

    The results of an experimental work aimed at improving the performances of the Cr-39 nuclear track detector for neutron dosimetry applications. So, a set of Cr-39 plastic detectors was exposed to 252 Cf neutron source, which has the emission rate of 0.68 x 10 8 s ( -1), and neutron dose equivalent rate 1m apart from the source is equal to 3.8 mrem/h. The detection of fast neutrons performed with Cr-39 detector foils, subsequent chemical etching and evaluation of the etched tracks by an automatic track counting system was studied. It is found that the track density grows with the increase of neutron dose and etching time. These results. are compared with previous work. It is found that there is a matching and good agreement with their investigations

  13. Advanced Neutron Source: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS

  14. Neutron generator (HIRRAC) and dosimetry study.

    Science.gov (United States)

    Endo, S; Hoshi, M; Takada, J; Tauchi, H; Matsuura, S; Takeoka, S; Kitagawa, K; Suga, S; Komatsu, K

    1999-12-01

    Dosimetry studies have been made for neutrons from a neutron generator at Hiroshima University (HIRRAC) which is designed for radiobiological research. Neutrons in an energy range from 0.07 to 2.7 MeV are available for biological irradiations. The produced neutron energies were measured and evaluated by a 3He-gas proportional counter. Energy spread was made certain to be small enough for radiobiological studies. Dose evaluations were performed by two different methods, namely use of tissue equivalent paired ionization chambers and activation of method with indium foils. Moreover, energy deposition spectra in small targets of tissue equivalent materials, so-called lineal energy spectrum, were also measured and are discussed. Specifications for biological irradiation are presented in terms of monoenergetic beam conditions, dose rates and deposited energy spectra.

  15. Neutron wave optics studied with ultracold neutrons

    International Nuclear Information System (INIS)

    Steyerl, A.

    1984-01-01

    The author discusses experiments demonstrating or utilizing the wave properties of neutrons with wavelengths of about 100 nm. In particular the 'UCN gravity diffractometer' and the gravity spectrometer NESSIE (Neutronen-Schwerkraft-Spectrometrie) are illustrated. (Auth.)

  16. A neutron source for IGISOL-JYFLTRAP: Design and characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Mattera, A.; Pomp, S.; Lantz, M.; Rakopoulos, V.; Solders, A.; Al-Adili, A.; Passoth, E.; Prokofiev, A.V.; Andersson, P.; Hjalmarsson, A. [Uppsala University, BOX 516, Uppsala (Sweden); Bedogni, R.; Esposito, A.; Gentile, A. [INFN-LNF, Frascati (Italy); Bortot, D. [INFN-LNF, Frascati (Italy); Politecnico di Milano, Milano (Italy); Gomez-Ros, J.M. [INFN-LNF, Frascati (Italy); CIEMAT, Madrid (Spain); Introini, M.V.; Pola, A. [Politecnico di Milano, Milano (Italy); Gorelov, D.; Penttilae, H.; Moore, I.D.; Rinta-Antila, S.; Kolhinen, V.S.; Eronen, T. [University of Jyvaeskylae (Finland)

    2017-08-15

    A white neutron source based on the Be(p, nx) reaction for fission studies at the IGISOL-JYFLTRAP facility has been designed and tested. 30MeV protons impinge on a 5mm thick water-cooled beryllium disc. The source was designed to produce at least 10{sup 12} fast neutrons/s on a secondary fission target, in order to reach competitive production rates of fission products far from the valley of stability. The Monte Carlo codes MCNPX and FLUKA were used in the design phase to simulate the neutron energy spectra. Two experiments to characterise the neutron field were performed: the first was carried out at The Svedberg Laboratory in Uppsala (SE), using an Extended-Range Bonner Sphere Spectrometer and a liquid scintillator which used the time-of-flight (TOF) method to determine the energy of the neutrons; the second employed Thin-Film Breakdown Counters for the measurement of the TOF, and activation foils, at the IGISOL facility in Jyvaeskylae (FI). Design considerations and the results of the two characterisation measurements are presented, providing benchmarks for the simulations. (orig.)

  17. Accelerator shield design of KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.

    2013-01-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

  18. Design of Double PG Crystal Neutron Diffractometer

    International Nuclear Information System (INIS)

    Adib, M.; Habib, N.; El-Mesiry, M.S.; Fathallah, M.

    2011-01-01

    The design of a diffractometer containing two pyrolytic graphite (PG) crystals to select monochromatic neutrons in the range of wavelengths longer than 0.26 nm is given. The first crystal is high oriented pyrolytic graphite (HOPG) set at glancing angle to reflect monochromatic neutrons with a selected wavelength. The second is a low quality PG crystal filter, set at take-off-angle such that, it transmits the selected monochromatic neutrons and rejects the higher order contaminations accompanying the first order reflection. It was shown that, 2 mm thick of PG crystal having 0.30 FWHM on mosaic spread are the optimum parameters of monochromator PG crystal. While the optimum thickness and mosaic spread of the PG crystal filter were selected to have low contamination factor of higher order reflections. The optimum parameters of the PG filter crystal were calculated using the computer package Graphite recently developed in our laboratory. Calculation shows that, 3 cm thick PG filter (20 on mosaic spread) is sufficient to almost eliminate the higher order contaminations accompanying the main monochromatic neutrons with

  19. Neutron spectroscopy for confinement studies

    International Nuclear Information System (INIS)

    Zorn, R.

    2010-01-01

    Neutron spectroscopy is an important method for the study of microscopic dynamics because it captures the spatial as well as the temporal aspects of the atomic or molecular motion. In this article techniques will be presented which are of special importance for the study of confined systems. Many of these are based on the fact that neutron scattering is isotope-dependent. Possible sources of systematic errors in measurements of confined systems will be pointed out. (author)

  20. Cold neutron source conceptual designing for Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khajvand, N.; Mirvakili, S.M.; Faghihi, F.

    2016-01-01

    Highlights: • Cold neutron source conceptual designing for Tehran research reactor is carried out. • Type and geometry of moderator and dimensions of cold neutron source are analyzed. • Liquid hydrogen with more ortho-concentration can be better option as moderator. - Abstract: A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options.

  1. Electron accelerator shielding design of KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

    2016-06-15

    The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

  2. Design innovations in neutron and gamma detectors

    International Nuclear Information System (INIS)

    Prasad, K.R.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  4. Design and simulation of an optimized e-linac based neutron source for BNCT research

    International Nuclear Information System (INIS)

    Durisi, E.; Alikaniotis, K.; Borla, O.; Bragato, F.; Costa, M.; Giannini, G.; Monti, V.; Visca, L.; Vivaldo, G.; Zanini, A.

    2015-01-01

    The paper is focused on the study of a novel photo-neutron source for BNCT preclinical research based on medical electron Linacs. Previous studies by the authors already demonstrated the possibility to obtain a mixed thermal and epithermal neutron flux of the order of 10"7 cm"−"2 s"−"1. This paper investigates possible Linac’s modifications and a new photo-converter design to rise the neutron flux above 5 10"7 cm"−"2 s"−"1, also reducing the gamma contamination. - Highlights: • Proposal of a mixed thermal and epithermal (named hyperthermal) neutron source based on medical high energy electron Linac. • Photo-neutron production via Giant Dipole Resonance on high Z materials. • MCNP4B-GN simulations to design the photo-converter geometry maximizing the hyperthermal neutron flux and minimizing the fast neutron and gamma contaminations. Hyperthermal neutron field suitable for BNCT preclinical research.

  5. Design of neutron radiography facility in pool for the reactor RA-10

    International Nuclear Information System (INIS)

    Peirone, M.; Coleff, A.; Sanchez, F.; Chiaraviglio, N.

    2013-01-01

    RA-10 project consists in the design and construction of a multipurpose reactor for multiple applications, including radioisotopes production, material testing and an in pool facility for neutron imaging. Neutron imaging is a powerful tool for studies of materials and offer several advantages among other attenuation-based techniques. In this study mechanical and neutronic requirements for the RA-10 in pool neutron imaging facility are described. The MCNP neutronic model and the mechanical design satisfying these requirements in a first engineering stage are described. (author)

  6. High Flux Isotope Reactor cold neutron source reference design concept

    International Nuclear Information System (INIS)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

  7. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  8. The Spallation Neutron Source accelerator system design

    Science.gov (United States)

    Henderson, S.; Abraham, W.; Aleksandrov, A.; Allen, C.; Alonso, J.; Anderson, D.; Arenius, D.; Arthur, T.; Assadi, S.; Ayers, J.; Bach, P.; Badea, V.; Battle, R.; Beebe-Wang, J.; Bergmann, B.; Bernardin, J.; Bhatia, T.; Billen, J.; Birke, T.; Bjorklund, E.; Blaskiewicz, M.; Blind, B.; Blokland, W.; Bookwalter, V.; Borovina, D.; Bowling, S.; Bradley, J.; Brantley, C.; Brennan, J.; Brodowski, J.; Brown, S.; Brown, R.; Bruce, D.; Bultman, N.; Cameron, P.; Campisi, I.; Casagrande, F.; Catalan-Lasheras, N.; Champion, M.; Champion, M.; Chen, Z.; Cheng, D.; Cho, Y.; Christensen, K.; Chu, C.; Cleaves, J.; Connolly, R.; Cote, T.; Cousineau, S.; Crandall, K.; Creel, J.; Crofford, M.; Cull, P.; Cutler, R.; Dabney, R.; Dalesio, L.; Daly, E.; Damm, R.; Danilov, V.; Davino, D.; Davis, K.; Dawson, C.; Day, L.; Deibele, C.; Delayen, J.; DeLong, J.; Demello, A.; DeVan, W.; Digennaro, R.; Dixon, K.; Dodson, G.; Doleans, M.; Doolittle, L.; Doss, J.; Drury, M.; Elliot, T.; Ellis, S.; Error, J.; Fazekas, J.; Fedotov, A.; Feng, P.; Fischer, J.; Fox, W.; Fuja, R.; Funk, W.; Galambos, J.; Ganni, V.; Garnett, R.; Geng, X.; Gentzlinger, R.; Giannella, M.; Gibson, P.; Gillis, R.; Gioia, J.; Gordon, J.; Gough, R.; Greer, J.; Gregory, W.; Gribble, R.; Grice, W.; Gurd, D.; Gurd, P.; Guthrie, A.; Hahn, H.; Hardek, T.; Hardekopf, R.; Harrison, J.; Hatfield, D.; He, P.; Hechler, M.; Heistermann, F.; Helus, S.; Hiatt, T.; Hicks, S.; Hill, J.; Hill, J.; Hoff, L.; Hoff, M.; Hogan, J.; Holding, M.; Holik, P.; Holmes, J.; Holtkamp, N.; Hovater, C.; Howell, M.; Hseuh, H.; Huhn, A.; Hunter, T.; Ilg, T.; Jackson, J.; Jain, A.; Jason, A.; Jeon, D.; Johnson, G.; Jones, A.; Joseph, S.; Justice, A.; Kang, Y.; Kasemir, K.; Keller, R.; Kersevan, R.; Kerstiens, D.; Kesselman, M.; Kim, S.; Kneisel, P.; Kravchuk, L.; Kuneli, T.; Kurennoy, S.; Kustom, R.; Kwon, S.; Ladd, P.; Lambiase, R.; Lee, Y. Y.; Leitner, M.; Leung, K.-N.; Lewis, S.; Liaw, C.; Lionberger, C.; Lo, C. C.; Long, C.; Ludewig, H.; Ludvig, J.; Luft, P.; Lynch, M.; Ma, H.; MacGill, R.; Macha, K.; Madre, B.; Mahler, G.; Mahoney, K.; Maines, J.; Mammosser, J.; Mann, T.; Marneris, I.; Marroquin, P.; Martineau, R.; Matsumoto, K.; McCarthy, M.; McChesney, C.; McGahern, W.; McGehee, P.; Meng, W.; Merz, B.; Meyer, R.; Meyer, R.; Miller, B.; Mitchell, R.; Mize, J.; Monroy, M.; Munro, J.; Murdoch, G.; Musson, J.; Nath, S.; Nelson, R.; Nelson, R.; O`Hara, J.; Olsen, D.; Oren, W.; Oshatz, D.; Owens, T.; Pai, C.; Papaphilippou, I.; Patterson, N.; Patterson, J.; Pearson, C.; Pelaia, T.; Pieck, M.; Piller, C.; Plawski, T.; Plum, M.; Pogge, J.; Power, J.; Powers, T.; Preble, J.; Prokop, M.; Pruyn, J.; Purcell, D.; Rank, J.; Raparia, D.; Ratti, A.; Reass, W.; Reece, K.; Rees, D.; Regan, A.; Regis, M.; Reijonen, J.; Rej, D.; Richards, D.; Richied, D.; Rode, C.; Rodriguez, W.; Rodriguez, M.; Rohlev, A.; Rose, C.; Roseberry, T.; Rowton, L.; Roybal, W.; Rust, K.; Salazer, G.; Sandberg, J.; Saunders, J.; Schenkel, T.; Schneider, W.; Schrage, D.; Schubert, J.; Severino, F.; Shafer, R.; Shea, T.; Shishlo, A.; Shoaee, H.; Sibley, C.; Sims, J.; Smee, S.; Smith, J.; Smith, K.; Spitz, R.; Staples, J.; Stein, P.; Stettler, M.; Stirbet, M.; Stockli, M.; Stone, W.; Stout, D.; Stovall, J.; Strelo, W.; Strong, H.; Sundelin, R.; Syversrud, D.; Szajbler, M.; Takeda, H.; Tallerico, P.; Tang, J.; Tanke, E.; Tepikian, S.; Thomae, R.; Thompson, D.; Thomson, D.; Thuot, M.; Treml, C.; Tsoupas, N.; Tuozzolo, J.; Tuzel, W.; Vassioutchenko, A.; Virostek, S.; Wallig, J.; Wanderer, P.; Wang, Y.; Wang, J. G.; Wangler, T.; Warren, D.; Wei, J.; Weiss, D.; Welton, R.; Weng, J.; Weng, W.-T.; Wezensky, M.; White, M.; Whitlatch, T.; Williams, D.; Williams, E.; Wilson, K.; Wiseman, M.; Wood, R.; Wright, P.; Wu, A.; Ybarrolaza, N.; Young, K.; Young, L.; Yourd, R.; Zachoszcz, A.; Zaltsman, A.; Zhang, S.; Zhang, W.; Zhang, Y.; Zhukov, A.

    2014-11-01

    The Spallation Neutron Source (SNS) was designed and constructed by a collaboration of six U.S. Department of Energy national laboratories. The SNS accelerator system consists of a 1 GeV linear accelerator and an accumulator ring providing 1.4 MW of proton beam power in microsecond-long beam pulses to a liquid mercury target for neutron production. The accelerator complex consists of a front-end negative hydrogen-ion injector system, an 87 MeV drift tube linear accelerator, a 186 MeV side-coupled linear accelerator, a 1 GeV superconducting linear accelerator, a 248-m circumference accumulator ring and associated beam transport lines. The accelerator complex is supported by ~100 high-power RF power systems, a 2 K cryogenic plant, ~400 DC and pulsed power supply systems, ~400 beam diagnostic devices and a distributed control system handling ~100,000 I/O signals. The beam dynamics design of the SNS accelerator is presented, as is the engineering design of the major accelerator subsystems.

  9. Recent Neutronic Optimization Studies at the SNS

    International Nuclear Information System (INIS)

    Murphy, B.D.; Ferguson, P.D.

    2002-01-01

    Recent design considerations at the Spallation Neutron Source have led to significant changes in the target station design, including changing the outer lead reflector to stainless steel and adding structural elements to aid heat transfer. In light of the design evolution, basic design decisions, including the moderator positions, were re-evaluated. With the proton beam energy of 1.0 GeV and a beam power of 2 MW, moderator positions were originally selected to optimize the performance of the upstream moderators, although some penalty was accepted in order to enhance the overall performance of the mixed coupled and decoupled moderators in the SNS target system. The work presented in this paper details sensitivity studies of selected moderator positions as a function of neutron energy. A possible change in proton beam energy, to 1.3 GeV while maintaining a beam power of 2 MW, is also studied in terms of moderator position. (authors)

  10. Mechanical design of epithermal neutron diagnostic for TFTR

    International Nuclear Information System (INIS)

    Groo, R.C.

    1981-01-01

    The mechanical design of the Epithermal Neutron Diagnostic for TFTR is described. This fission detector system measures the time resolution of the neutron flux for folding into the Neutron Activation system and also provides continuous, wide range coverage of all expected fusion reaction rates

  11. AHTR Mechanical, Structural, and Neutronic Preconceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Varma, V.K.; Holcomb, D.E.; Peretz, F.J.; Bradley, E.C.; Ilas, D.; Qualls, A.L.; Zaharia, N.M.

    2012-09-15

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design

  12. AHTR Mechanical, Structural, And Neutronic Preconceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Varma, Venugopal Koikal [ORNL; Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Bradley, Eric Craig [ORNL; Ilas, Dan [ORNL; Qualls, A L [ORNL; Zaharia, Nathaniel M [ORNL

    2012-10-01

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic

  13. Basic Design of the Cold Neutron Research Facility in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, K. H.; Kim, Y. K.

    2005-09-01

    The HANARO Cold Neutron Research Facility (CNRF) Project has been embarked in July 2003. The CNRF project has selected as one of the radiation technology development project by National Science and Technology Committee in June 2002. In this report, the output of the second project year is summarized as a basic design of cold neutron source and related systems, neutron guide, and neutron scattering instruments

  14. Basic Design of the Cold Neutron Research Facility in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hark Rho; Lee, K. H.; Kim, Y. K. (and others)

    2005-09-15

    The HANARO Cold Neutron Research Facility (CNRF) Project has been embarked in July 2003. The CNRF project has selected as one of the radiation technology development project by National Science and Technology Committee in June 2002. In this report, the output of the second project year is summarized as a basic design of cold neutron source and related systems, neutron guide, and neutron scattering instruments.

  15. Study of the RP-10 reactor neutron beam applied to the neutron radiography

    International Nuclear Information System (INIS)

    Zegarra, Manuel; Lopez, Alcides

    2013-01-01

    We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 10 6 and 10 4 n/cm 2 s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

  16. HFIR cold neutron source moderator vessel design analysis

    International Nuclear Information System (INIS)

    Chang, S.J.

    1998-04-01

    A cold neutron source capsule made of aluminum alloy is to be installed and located at the tip of one of the neutron beam tubes of the High Flux Isotope Reactor. Cold hydrogen liquid of temperature approximately 20 degree Kelvin and 15 bars pressure is designed to flow through the aluminum capsule that serves to chill and to moderate the incoming neutrons produced from the reactor core. The cold and low energy neutrons thus produced will be used as cold neutron sources for the diffraction experiments. The structural design calculation for the aluminum capsule is reported in this paper

  17. Neutron diffraction studies of glasses

    International Nuclear Information System (INIS)

    Wright, A.C.

    1987-01-01

    A survey is given of the application of neutron diffraction to structural studies of oxide and halide glasses. As with crystalline materials, neutron and X-ray diffraction are the major structural probes for glasses and other amorphous solids, particularly in respect of intermediate range order. The glasses discussed mostly have structures which are dominated by a network in which the bonding is predominantly covalent. The examples discussed demonstrate the power of the neutron diffraction technique in the investigation of the structures of inorganic glasses. The best modern diffraction experiments are capable of providing accurate data with high real space resolution, which if used correctly, are an extremely fine filter for the various structural models proposed in the literature. 42 refs

  18. The advanced neutron source design - A status report

    International Nuclear Information System (INIS)

    West, C.D.

    1992-01-01

    The Advanced Neutron Source (ANS) facility is being designed as a user laboratory for all types of neutron-based research, centered around a nuclear fission reactor (D 2 O cooled, moderated, and reflected), operating at approximately 300 MWth. Safety, and especially passive safety features, have been emphasized throughout the design process. The design also provides experimental facilities for neutron scattering and nuclear and fundamental physics research, transuranic and other isotope production, radiation effects research, and materials analysis. (author)

  19. Design, construction and characterization of a dosimeter for neutron radiation

    International Nuclear Information System (INIS)

    Souto, Eduardo de Brito

    2007-01-01

    An individual dosimeter for neutron-gamma mixed field dosimetry was design and developed aiming monitoring the increasing number of workers potentially exposed to neutrons. The proposed dosimeter was characterized to an Americium-Beryllium source spectrum and dose range of radiation protection interest (up to 20 mSv). Thermoluminescent albedo dosimetry and nuclear tracks dosimetry, traditional techniques found in the international literature, with materials of low cost and national production, were used. A commercial polycarbonate, named SS-1, was characterized for solid state tack detector application. The chemical etching parameters and the methodology of detectors evaluation were determined. The response of TLD-600, TLD-700 and SS-1 were studied and algorithms for dose calculation of neutron and gamma radiation of Americium- Beryllium sources were proposed. The ratio between thermal, albedo and fast neutrons responses, allows analyzing the spectrum to which the dosimeter was submitted and correcting the track detector response to variations in the radiation incidence angle. The new dosimeter is fully characterized, having sufficient performance to be applied as neutron dosimeter in Brazil. (author)

  20. Experimental study of angular neutron flux spectra on a slab surface to assess nuclear data and calculational methods for a fusion reactor design

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1988-06-01

    This paper presents an experimental approach to interpret the results of integral experiments for fusion neutronics research. The measurement is described of the angular neutron flux on a restricted area of slab assemblies with D-T neutron bombardment by using the time-of-flight (TOF) method with an NE213 neutron detector over an energy range from 0.05 to 15 MeV. A two bias scheme was developed to obtain an accurate detection efficiency over a wide energy range. The detector-collimator response function was introduced to define the restricted surface area and to determine the effective measured area. A series of measurements of the angular neutron flux on slabs of fusion blanket materials, i.e., Be, C, and Li 2 O, as functions of neutron leaking angle and slab thickness have been performed to examine neutron transport characteristics in bulk materials. The calculational analyses of the experimental results have been also carried out by using Monte Carlo neutron transport codes, i.e., MORSE-DD and MCNP. The existing nuclear data files, i.e., JENDL-3PR1, -3PR2, ENDF/B-IV and -V were tested by comparing with the experimental results. From the comparisons, the data on C and 7 Li in the present files are fairly sufficient. Those on beryllium, however, is insufficient for the estimation of high threshold reactions such as tritium production in a fusion reactor blanket design. It is also found that the total and elastic cross sections are more important for accurate predictions of neutronic parameters at deep position. The comparisons between the measured and calculated results provide information to understand the results of the previous integral experiments for confirmation of accuracy of fusion reactor designs. (author)

  1. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  2. A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

    International Nuclear Information System (INIS)

    Choi, H.; Downar, T.J.

    1999-01-01

    A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

  3. An experimental facility for studying delayed neutron emission

    International Nuclear Information System (INIS)

    Dermendzhiev, E.; Nazarov, V.M.; Pavlov, S.S.; Ruskov, Iv.; Zamyatin, Yu.S.

    1993-01-01

    A new experimental facility for studying delayed neutron emission has been designed and tested. A method based on utilization of the Dubna IBR-2 pulsed reactor, has been proposed and realized for periodical irradiation of targets composed of fissionable isotopes. Such a powerful pulsed neutron source in combination with a slow neutron chopper synchronized with the reactor bursts makes possible variation of the exposure duration and effective suppression of the fast neutron background due to delay neutrons emitted from the reactor core. Detection of delayed neutrons from the target is carried out by a high-efficiency multicounter neutron detector with a near-4π geometry. Some test measurements and results are briefly described. Possible use of the facility for other tasks is also discussed. 14 refs.; 14 figs

  4. Spallation neutron source target station design, development, and commissioning

    Energy Technology Data Exchange (ETDEWEB)

    Haines, J.R., E-mail: hainesjr@ornl.gov; McManamy, T.J.; Gabriel, T.A.; Battle, R.E.; Chipley, K.K.; Crabtree, J.A.; Jacobs, L.L.; Lousteau, D.C.; Rennich, M.J.; Riemer, B.W.

    2014-11-11

    The spallation neutron source target station is designed to safely, reliably, and efficiently convert a 1 GeV beam of protons to a high flux of about 1 meV neutrons that are available at 24 neutron scattering instrument beam lines. Research and development findings, design requirements, design description, initial checkout testing, and results from early operation with beam are discussed for each of the primary target subsystems, including the mercury target, neutron moderators and reflector, surrounding vessels and shielding, utilities, remote handling equipment, and instrumentation and controls. Future plans for the mercury target development program are also briefly discussed.

  5. Optimization study and neutronic and thermal-hydraulic design calculations of a 75 KWTH aqueous homogeneous reactor for medical isotopes production

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Garcia, Lorena P. Rodriguez; Llanes, Jesus Salomon; Hernandez, Carlos R. Garcia, E-mail: dperez@instec.cu, E-mail: dmilian@instec.cu, E-mail: lorenapilar@instec.cu, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Lira, Carlos A. Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife (Brazil); Rodriguez, Manuel Cadavid, E-mail: mcadavid2001@yahoo.com [Tecnologia Nuclear Medica Spa, TNM (Chile)

    2015-07-01

    {sup 99m}Tc is the most common radioisotope used in nuclear medicine. It is a very useful radioisotope, which is used in about 30-40 million procedures worldwide every year. Medical diagnostic imaging techniques using {sup 99m}Tc represent approximately 80% of all nuclear medicine procedures. Although {sup 99m}Tc can be produced directly on a cyclotron or other type of particle accelerator, currently is almost exclusively produced from the beta-decay of its 66-h parent {sup 99}Mo. {sup 99}Mo production system in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. In this paper, an AHR conceptual design using Low Enriched Uranium (LEU) is studied and optimized for the production of {sup 99}Mo. Aspects related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotopes production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution due to thermal expansion of the fuel solution and the void volume generated by radiolytic gas bubbles were evaluated. Thermal-hydraulics studies were carried out in order to show that sufficient cooling capacity exists to prevent fuel overheating. The neutronic and thermal-hydraulics calculations have been performed with the MCNPX computational code and the version 14 of ANSYS CFX respectively. The neutronic calculations demonstrated that the reactor is able to produce 370 six-day curies of {sup 99}Mo in 5 days operation cycles and the CFD simulation demonstrated that the heat removal systems provide sufficient cooling capacity to prevent fuel overheating, the maximum temperature reached by the fuel (89.29 deg C) was smaller to the allowable temperature limit (90 deg C). (author)

  6. Advanced Neutron Source radiological design criteria

    International Nuclear Information System (INIS)

    Westbrook, J.L.

    1995-08-01

    The operation of the proposed Advanced Neutron Source (ANS) facility will present a variety of radiological protection problems. Because it is desired to design and operate the ANS according to the applicable licensing standards of the Nuclear Regulatory Commission (NRC), it must be demonstrated that the ANS radiological design basis is consistent not only with state and Department of Energy (DOE) and other usual federal regulations, but also, so far as is practicable, with NRC regulations and with recommendations of such organizations as the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI). Also, the ANS radiological design basis is in general to be consistent with the recommendations of authoritative professional and scientific organizations, specifically the National Council on Radiation Protection and Measurements (NCRP) and the International Commission on Radiological Protection (ICRP). As regards radiological protection, the principal goals of DOE regulations and guidance are to keep occupational doses ALARA [as low as (is) reasonably achievable], given the current state of technology, costs, and operations requirements; to control and monitor contained and released radioactivity during normal operation to keep public doses and releases to the environment ALARA; and to limit doses to workers and the public during accident conditions. Meeting these general design objectives requires that principles of dose reduction and of radioactivity control by employed in the design, operation, modification, and decommissioning of the ANS. The purpose of this document is to provide basic radiological criteria for incorporating these principles into the design of the ANS. Operations, modification, and decommissioning will be covered only as they are affected by design

  7. Neutron reflector design with Californium 252 neutron for Boron neutron chapter therapy facility using MCNP5 simulation method

    International Nuclear Information System (INIS)

    Muhammad Fakhrurreza; Kusminanto; Y Sardjono

    2014-01-01

    In this research has made a reflector design to provide beams of Neutron for BNCT with Californium-252 radioactive source. This collimator is useful to obtain optimum epithermal neutron flux with the smallest impurity radiation (thermal neutron, fast neutron, and gamma). The design process is done using Monte Carlo N-Particle simulation version 5 (MCNP5) code to calculate the neutron flux tally form. The chosen reflector design is the reflectors which use material such as BeO ceramic with 13 cm thick. Moderator use sulfur material with the slope angle of the cone is 30°. From the calculation result, it is obtained that Reflector with 1 gram Californium-252 source can produce a neutron output thermal which has thermal neutron specification 2.23189 x 10 9 n/s.cm 2 , epithermal neutron 3.51548 x 10 9 n/s.cm 2 , and fast neutron 4.82241 x 10 9 n/s.cm 2 From the result, it needs additional collimator because the BNCT requirement. (author)

  8. Neutron beam studies

    International Nuclear Information System (INIS)

    Willis, B.T.M.; Apps, M.E.

    1979-01-01

    The report is in sections, entitled: metallurgy; reactor fuels; defect solid state and interatomic forces; surface studies; colloid, polymer and biological studies; glasses; energy-related materials; solid state physics and crystallography; instrumentation and experimental technique. (U.K.)

  9. A pulsed source neutron reflectometer for surface studies

    International Nuclear Information System (INIS)

    Penfold, J.; Williams, W.G.

    1985-05-01

    A design for a neutron reflectometer for surface studies to be constructed at the SNS is presented. Examples of its use to study problems in surface chemistry, surface magnetism and low dimensional structures are highlighted. (author)

  10. Design, Construction, and Modeling of a 252Cf Neutron Irradiator

    Directory of Open Access Journals (Sweden)

    Blake C. Anderson

    2016-01-01

    Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

  11. Design and characterisation of a pulsed neutron interrogation facility

    International Nuclear Information System (INIS)

    Favalli, A.; Pedersen, B.

    2007-01-01

    The Joint Research Centre recently obtained a license to operate a new experimental device intended for research in the field of nuclear safeguards. The research projects currently being planned for the new device includes mass determination of fissile materials in matrices and detection of contraband non-nuclear materials. The device incorporates a commercial pulsed neutron generator and a large graphite mantle surrounding the sample cavity. In this configuration, a relatively high thermal neutron flux with a long lifetime is achieved inside the sample cavity. By pulsing the neutron generator, a sample may be interrogated by a pure thermal neutron flux during repeated time periods. The paper reports on the design of the new device and the pulsed fast and thermal neutron source. The thermal neutron flux caused by the neutron generator and the graphite structure has been characterised by foil activation, fission chamber and 3 He proportional counter measurements. (authors)

  12. Influence of instrument design on neutron lifetime measurements

    International Nuclear Information System (INIS)

    Youmans, A.H.; Hopkinson, E.C.

    1975-01-01

    Commercially available logging services provide a measurement of the lifetime of thermal neutrons in formations adjacent to a borehole. This lifetime provides a measure of the macroscopic thermal neutron-capture cross-section Σ of the formation, which in turn is functionally related to the abundance and constituency of the rock matrix and contained fluids. Because the measurement is extremely sensitive to an abundance of trace elements like boron and gadolinium, it is very difficult to find rock formations with an accurately known value of Σ, which is required for the accuracy of the measuring system to be experimentally tested. Various theoretical studies published suggest that errors in the determination of Σ may occur because of the influence of borehole parameters and the effects of neutron diffusion. Experimental results are reported that demonstrate that the design of the instrument is crucial to the validity of any theoretical treatment of the subject. The influence of neutron diffusion and borehole effects can be overcome by optimal selection of spacing and shielding parameters

  13. Neutron scattering studies in the actinide region

    International Nuclear Information System (INIS)

    Beghian, L.E.; Kegel, G.H.R.

    1991-08-01

    During the report period we have investigated the following areas: Neutron elastic and inelastic scattering measurements on 14 N, 181 Ta, 232 Th, 238 U and 239 Pu; Prompt fission spectra for 232 Th, 235 U, 238 U and 239 Pu; Theoretical studies of neutron scattering; Neutron filters; New detector systems; and Upgrading of neutron target assembly, data acquisition system, and accelerator/beam-line apparatus

  14. Recent development in magnetic neutron scattering studies

    International Nuclear Information System (INIS)

    Endoh, Yasuo

    1993-01-01

    Neutron scattering results contain many new concepts in modern magnetism. We review here the most recent neutron magnetic scattering studies from so called '214' copper oxide lamellar materials, because a number of important developments in magnetism are condensed in this novel subject. We show that neutron scattering has played crucial role in our understanding of modern magnetism. (author)

  15. Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

    International Nuclear Information System (INIS)

    Shaaban, I.

    2009-11-01

    Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

  16. The design and analysis of integral assembly experiments for CTR neutronics

    International Nuclear Information System (INIS)

    Beynon, T.D.; Curtis, R.H.; Lambert, C.

    1978-01-01

    The use of simple-geometry integral assemblies of lithium metal or lithium compounds for the study of the neutronics of various CTR designs is considered and four recent experiments are analysed. The relatively long mean free path of neutrons in these assemblies produces significantly different design problems from those encountered in similar experiments for fission reactor design. By considering sensitivity profiles for various parameters it is suggested that experiments can be designed to be optimised for data adjustments. (author)

  17. 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

  18. Study of a Loop Heat Pipe Using Neutron Radiography

    International Nuclear Information System (INIS)

    C. Thomas Conroy; A. A. El-Ganayni; David R. Riley; John M. Cimbala; Jack S. Brenizer, Jr.; Abel Po-Ya Chuang; Shane Hanna

    2001-01-01

    An explanation is given of what a loop heat pipe (LHP) is, and how it works. It is then shown that neutron imaging (both real time neutron radioscopy and single exposure neutron radiography) is an effective experimental tool for the study of LHPs. Specifically, neutron imaging has helped to identify and correct a cooling water distribution problem in the condenser, and has enabled visualization of two-phase flow (liquid and vapor) in various components of the LHP. In addition, partial wick dry-out, a phenomenon of great importance in the effective operation of LHPs, has been identified with neutron imaging. It is anticipated that neutron radioscopy and radiography will greatly contribute to our understanding of LHP operation, and will lead to improvement of LHP modeling and design

  19. Neutronics design for a spherical tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

    Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

  20. Methods for the neutronic design of a Supersara experimental loop

    International Nuclear Information System (INIS)

    Casali, F.; Cepraga, D.

    1982-01-01

    This paper describes a method for the neutronic design of experimental loops irradiated in D 2 O experimental reactors, like Essor. The calculation approach concerns the definition of a Weigner-Seitz cell where the loop under examination be subjected to the same neutronic conditions as in the actual reactor

  1. Study of neutron focusing at the Texas Cold Neutron Source. Final report

    International Nuclear Information System (INIS)

    Wehring, B.W.; Uenlue, K.

    1995-01-01

    Funds were received for the first year of a three year DOE Nuclear Engineering Research Grant, ''Study of Neutron Focusing at the Texas Cold Neutron Source'' (FGO2-92ER75711). The purpose of this three year study was to develop a neutron focusing system to be used with the Texas Cold Neutron Source (TCNS) to produce an intense beam of neutrons. A prompt gamma activation analysis (PGAA) facility was also to be designed, setup, and tested under the three year project. During the first year of the DOE grant, a new procedure was developed and used to design a focusing converging guide consisting of truncated rectangular cone sections. Detailed calculations were performed using a 3-D Monte Carlo code which we wrote to trace neutrons through the curved guide of the TCNS into the proposed converging guide. Using realistic reflectivities for Ni-Ti supermirrors, we obtained gains of 3 to 5 for the neutron flux averaged over an area of 1 x 1 cm

  2. A Dosimetry Study of Deuterium-Deuterium Neutron Generator-based In Vivo Neutron Activation Analysis.

    Science.gov (United States)

    Sowers, Daniel; Liu, Yingzi; Mostafaei, Farshad; Blake, Scott; Nie, Linda H

    2015-12-01

    A neutron irradiation cavity for in vivo neutron activation analysis (IVNAA) to detect manganese, aluminum, and other potentially toxic elements in human hand bone has been designed and its dosimetric specifications measured. The neutron source is a customized deuterium-deuterium neutron generator that produces neutrons at 2.45 MeV by the fusion reaction 2H(d, n)3He at a calculated flux of 7 × 10(8) ± 30% s(-1). A moderator/reflector/shielding [5 cm high density polyethylene (HDPE), 5.3 cm graphite and 5.7 cm borated (HDPE)] assembly has been designed and built to maximize the thermal neutron flux inside the hand irradiation cavity and to reduce the extremity dose and effective dose to the human subject. Lead sheets are used to attenuate bremsstrahlung x rays and activation gammas. A Monte Carlo simulation (MCNP6) was used to model the system and calculate extremity dose. The extremity dose was measured with neutron and photon sensitive film badges and Fuji electronic pocket dosimeters (EPD). The neutron ambient dose outside the shielding was measured by Fuji NSN3, and the photon dose was measured by a Bicron MicroREM scintillator. Neutron extremity dose was calculated to be 32.3 mSv using MCNP6 simulations given a 10-min IVNAA measurement of manganese. Measurements by EPD and film badge indicate hand dose to be 31.7 ± 0.8 mSv for neutrons and 4.2 ± 0.2 mSv for photons for 10 min; whole body effective dose was calculated conservatively to be 0.052 mSv. Experimental values closely match values obtained from MCNP6 simulations. These are acceptable doses to apply the technology for a manganese toxicity study in a human population.

  3. Pulsed neutron source cold moderators --- concepts, design and engineering

    International Nuclear Information System (INIS)

    Bauer, Guenter S.

    1997-01-01

    Moderator design for pulsed neutron sources is becoming more and more an interface area between source designers and instrument designers. Although there exists a high degree of flexibility, there are also physical and technical limitations. This paper aims at pointing out these limitations and examining ways to extend the current state of moderator technology in order to make the next generation neutron sources even more versatile and flexible tools for science in accordance with the users' requirements. (auth)

  4. Studies and modeling of cold neutron sources

    International Nuclear Information System (INIS)

    Campioni, G.

    2004-11-01

    With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources

  5. A shielding design for an accelerator-based neutron source for boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Hawk, A.E.; Blue, T.E. E-mail: blue.1@osu.edu; Woollard, J.E

    2004-11-01

    Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a {sup 7}Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.

  6. Design characteristics of a three-component AEOI Neutriran Albedo Neutron Personnel Dosimeter

    International Nuclear Information System (INIS)

    Sohrabi, M.; Katouzi, M.

    1991-01-01

    In establishing a national personnel neutron dosimetry service in Iran, different parameters of the AEOI Neutriran Albedo Neutron Personnel Dosimeter (NANPD) have been optimized. A NANPD was designed with three dosimetry components to measure (a) direct thermal neutrons, (b) direct fast neutrons and (C) direct neutrons by the detection of the albedo neutrons reflected from the body. The dosimeter consists of one or more Lexan polycarbonate and/or CR-39 foils and two 10 B (n,α) 7 Li converters in a cadmium cover so arranged as to efficiently measure the three neutron dose components separately. The boron converter thickness, its position relative to the beam direction and its distance from the PC foil were studied and the results were incorporated into the design. The dose response of the dosimeter, its lower detection limit as well as the correction factors related to the field neutrons and albedo neutrons were also determined for a 238 Pu-Be, an 241 Am-Be and a 252 Cf sources. In this paper, the dosimeter design and its dosimetric characteristics are presented and discussed. (author)

  7. Design of small ECR ion source for neutron generator

    International Nuclear Information System (INIS)

    Zhou Changgeng; Lou Benchao; Zu Xiulan; Yang Haisu; Xiong Riheng

    2003-01-01

    The principles, structures and characteristics of small ECR (Electron Cyclotron Resonance) ion source used in the neutron generator are introduced. The processes of the design and key technique and innovations are described. (authors)

  8. Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

    2015-07-01

    In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

  9. Study of spectral response of a neutron filter. Design of a method to adjust spectra; Etude des moyens de conditionnement de la reponse spectrale d'un filtre a neutrons. Mise au point d'une methode d'ajustement rapide de spectre

    Energy Technology Data Exchange (ETDEWEB)

    Colomb-Dolci, F. [Universite Louis Pasteur, 67 - Strasbourg (France)

    1999-02-01

    The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

  10. Design of filtered epithermal neutron beams for BNC

    International Nuclear Information System (INIS)

    Greenwood, R.C.

    1986-01-01

    The design principles of filters (installed in nuclear reactors) to provide epithermal neutron beams suitable for use in 10 B Neutron Capture Therapy (BNCT) are reviewed. The goal of such filters is to provide epithermal neutron beams within an energy range of 1 keV to 30 keV with fluxes in excess of 5 x 10 8 neutrons/cm 2 .s, and having acceptably low contaminant fast neutron (> 30 keV) and gamma components. Filters considered for this application include 238 U, Sc, Fe/Al and Al/S. It is shown that in order to achieve a goal epithermal neutron flux of > 5 x 10 8 neutrons/cm 2 .s, such filters must be located in radial beam channels which view essentially the complete reactor core. Based on considerations of estimated epithermal fluxes, cost and availability of materials, and transmitted neutron energy spectrum, it is suggested that a filter consisting of elements of Al, S, Ti and V might prove to be an optimum design for BNCT applications. 13 references, 3 figures, 8 tables

  11. Design of a system for neutrons dosimetry

    International Nuclear Information System (INIS)

    Ceron, P.; Rivera, T.; Paredes G, L.; Azorin, J.; Sanchez, A.; Vega C, H. R.

    2014-08-01

    At the present time diverse systems of detection of neutrons exist, as proportional counters based on BF 3 , He 3 and spectrometers of Bonner spheres. However, the cost and the complexity of the implementation of these systems put them far from the reach for dosimetric purposes. For these reasons a system of neutrons detection composed by a medium paraffin moderator that forms a 4π (spheres) arrangement and of several couples of thermoluminescent dosimeters TLD 600/TLD 700. The response of the system presents a minor repeatability to 5% in several assays when being irradiated with a 239 PuBe source and a deviation of 13.8% in the Tl readings of four different spheres. The calibration factor of the system with regard to the neutrons source which was of 56.2 p Sv/nc also was calculated. These detectors will be used as passive monitors of photoneutrons in a radiotherapy room with lineal accelerator of high energy. (Author)

  12. Californium Multiplier. Part I. Design for neutron radiography

    International Nuclear Information System (INIS)

    Crosbie, K.L.; Preskitt, C.A.; John, J.; Hastings, J.D.

    1982-01-01

    The Californium Multiplier (CFX) is a subcritical assembly of enriched uranium surrounding a californium-252 neutron source. The function of the CFX is to multiply the neutrons emitted by the source to a number sufficient for neutron radiography. The CFX is designed to provide a collimated beam of thermal neutrons from which the gamma radiation is filtered, and the scattered neutrons are reduced to make it suitable for high resolution radiography. The entire system has inherent safety features, which provide for system and personnel safety, and it operates at moderate cost. In Part I, the CFX and the theory of its operation are described in detail. Part II covers the performance of the Mound Facility CFX

  13. Study of calculated and measured time dependent delayed neutron yields

    International Nuclear Information System (INIS)

    Waldo, R.W.

    1980-05-01

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232 U, 237 Np, 238 Pu, 241 Am, /sup 242m/Am, 245 Cm, and 249 Cf were studied for the first time. The delayed neutron emission from 232 Th, 233 U, 235 U, 238 U, 239 Pu, 241 Pu, and 242 Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232 Th to 252 Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables

  14. Design of A HPGe-Plastic Scintillator Compton Suppression Spectrometer for Neutron Activation Analysis and Radio environmental Studies

    International Nuclear Information System (INIS)

    Sharshar, T.; Badran, H.; Elnimr, T.

    1998-01-01

    The design of a compton suppression spectrometer consisting of a 10% p-type HPGe detector and an annular anti-Compton shield made of N E-102 A plastic scintillator is described. The height of the guard plastic- scintillation detector was optimized experimentally using a NaI(Ti) ring, consisting of five NaI(Ti) detectors. The annular guard detector is divided to four optically isolated quarters to enhance the light collection. Each quarter of the guard detector was tested and satisfying results are obtained

  15. Studies of neutron measurement methods for fusion plasma diagnostics

    International Nuclear Information System (INIS)

    Beimer, K.H.

    1986-03-01

    This thesis comprises several studies mainly devoted to neutron measurement systems for plasma diagnostics at JET (Joint European Torus). An in situ calibration of the U-235 fission chamber detectors located at JET is presented. These detectors are used for measuring the neutron yield from the thermonuclear reactions in the plasma. The energy spectrum of the neutrons from the reactions D(d,n) 3 He has been studied by means of a 3 He spectrometer. Especially, it was found that by measuring the width of the full energy peak in the response spectrum of the 3 He-spectrometer, the deuterium distribution in the deuterium targets used can be estimated. In order to measure different neutron energies it is necessary to obtain a detailed knowledge of the response of the spectrometer. Therefore, the response function to monoenergetic neutrons in the energy range 130-3030 keV was experimentally determined. Some work has been related to a design study of a 14 MeV spectrometer for neutron diagnostics. It is a combined proton-recoil and time-of-flight spectrometer for high resolution measurements. The main parts of it are the collimator, the scattering foil, and the detectors for the recoil protons and the scattered neutrons. The influence of proton straggling in the foil on the resolution and efficiency of the spectrometer has been studied. Furthermore, a three dimensional Monte Carlo code has been written and used for the design of the collimator. (author)

  16. Gas target neutron generator studies

    International Nuclear Information System (INIS)

    Chatoorgoon, V.

    1978-01-01

    The need for an intense neutron source for the study of radiation damage on materials has resulted in the proposal of various solid, liquid, and gas targets. Among the gas targets proposed have been the transonic gas target, two types of hypersonic gas target, and the subsonic gas target (SGT). It has been suggested that heat deposition in a subsonic channel might create a gas density step which would constitute an attractive gas target type. The first part of the present study examines this aspect of the SGT and shows that gas density gradients are indeed formed by heat deposition in subsonic flow. The variation of beam voltage, gas density, gas pressure, and gas temperature within the channel have been calculated as functions of the system parameters: beam voltage, beam current, channel diameter, stagnation tank temperature and pressure. The analysis is applicable to any beam particle and target gas. For the case of T + on D 2 , which is relevant to the fusion application, the 14 MeV neutron profiles are presented as a function of system parameters. It is found that the SGT is compatible with concentrated intense source operation. The possibility of instability was investigated in detail using a non-linear analysis which made it possible to follow the complete time development of the SGT. It was found that the SGT is stable against all small perturbations and certain types of large perturbations. It appears that the SGT is the most advantageous type of gas target, operating at a lower mass flow and less severe stagnation tank conditions than the other types. The second part of the thesis examines a problem associated with the straight hypersonic target, the deuterium spill into the tritium port. The regime of practical operation for this target is established. (auth)

  17. Radiation shielding design of BNCT treatment room for D-T neutron source.

    Science.gov (United States)

    Pouryavi, Mehdi; Farhad Masoudi, S; Rahmani, Faezeh

    2015-05-01

    Recent studies have shown that D-T neutron generator can be used as a proper neutron source for Boron Neutron Capture Therapy (BNCT) of deep-seated brain tumors. In this paper, radiation shielding calculations have been conducted based on the computational method for designing a BNCT treatment room for a recent proposed D-T neutron source. By using the MCNP-4C code, the geometry of the treatment room has been designed and optimized in such a way that the equivalent dose rate out of the treatment room to be less than 0.5μSv/h for uncontrolled areas. The treatment room contains walls, monitoring window, maze and entrance door. According to the radiation protection viewpoint, dose rate results of out of the proposed room showed that using D-T neutron source for BNCT is safe. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. 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)

  19. Design considerations for neutron activation and neutron source strength monitors for ITER

    International Nuclear Information System (INIS)

    Barnes, C.W.; Jassby, D.L.; LeMunyan, G.; Roquemore, A.L.

    1997-01-01

    The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with ∼1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system

  20. Layered magnets: polarized neutron reflection studies

    Energy Technology Data Exchange (ETDEWEB)

    Zabel, H; Schreyer, A [Ruhr-Univ. Bochum, Lehrstuhl fuer Experimentalphysik/Festkoerperphysik, Bochum (Germany)

    1996-11-01

    Neutron reflectivity measurements from extended surfaces, thin films and superlattices provide information on the chemical profile parallel to the film normal, including film thicknesses, average composition and interfacial roughness parameters. Reflectivity measurements with polarized neutrons are particularly powerful for analyzing the magnetic density profiles in thin films and superlattices in addition to chemical profiles. The basic theory of polarized neutron reflectivity is provided, followed by some examples and more recent applications concerning polarized neutron reflectivity studies from exchange coupled Fe/Cr superlattices. (author) 5 figs., 13 refs.

  1. Target system neutronics study for NXGENS

    International Nuclear Information System (INIS)

    Willis, C.; Muhrer, G.

    2007-01-01

    The Materials Test Station (MTS) [E. Pitcher, G. Muhrer, H. Trellue, Neutronics Assessment of the LANSCE Materials Test Station as an Irradiation Facility for the JIMO Space Reactor, LA-CP-04-0903.], a spallation target station, planned for construction at the Los Alamos Neutron Science Center (LANSCE), will provide the opportunity to test the prototype of a long-pulse spallation source neutron scattering instrument (NXGENS). In this paper, we present the target-moderator neutronics optimization study that was performed in support of NXGENS

  2. Design and safety aspects of the Cornell cold neutron source

    International Nuclear Information System (INIS)

    Ouellet, Carol G.; Clark, David D.

    1992-01-01

    The cold neutron beam facility at the Cornell University TRIGA Mark II reactor will begin operational testing in early 1993. It is designed to provide a low background subthermal neutron beam that is as free as possible of fast neutrons and gamma rays for applied research and graduate-level instruction. The Cornell cold neutron source differs from the more conventional types of cold sources in that it is inherently safer because it uses a safe handling material (mesitylene) as the moderator instead of hydrogen or methane, avoids the circulation of cryogenic fluids by removing heat from the system by conduction through a 99.99% pure copper rod attached to a cryogenic refrigerator, and is much smaller in its size and loads. The design details and potential hazards are described, where it is concluded that no credible accident involving the cold source could cause damage to the reactor or personnel, or cause release of radioactivity. (author)

  3. Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

    International Nuclear Information System (INIS)

    Thiagu Supramaniam

    2007-01-01

    The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

  4. Collimator design for neutron imaging of laser-fusion targets

    International Nuclear Information System (INIS)

    Sommargren, G.E.; Lerche, R.A.

    1981-01-01

    Several pinhole collimator geometries for use in neutron imaging experiments have been modeled and compared. Point spread functions are shown for a cylinder, hyperbola, intersecting cones, and a five-zone approximation to the intersecting cones. Of the geometries studied, the intersecting cones appear the most promising with respect to neutron efficiency, field of view, and isoplanatism

  5. The Neutron star Interior Composition Explorer (NICER): design and development

    OpenAIRE

    Gendreau, Keith C.; Arzoumanian, Zaven; Adkins, Phillip W.; Albert, Cheryl L.; Anders, John F.; Aylward, Andrew T.; Baker, Charles L.; Balsamo, Erin R.; Bamford, William A.; Benegalrao, Suyog S.; Berry, Daniel L.; Bhalwani, Shiraz; Black, J. Kevin; Blaurock, Carl; Bronke, Ginger M.

    2016-01-01

    During 2014 and 2015, NASA's Neutron star Interior Composition Explorer (NICER) mission proceeded successfully through Phase C, Design and Development. An X-ray (0.2-12 keV) astrophysics payload destined for the International Space Station, NICER is manifested for launch in early 2017 on the Commercial Resupply Services SpaceX-11 flight. Its scientific objectives are to investigate the internal structure, dynamics, and energetics of neutron stars, the densest objects in the universe. During P...

  6. Design of neutron 'fluse-trop' in Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khalafi, H.

    1999-01-01

    In this project, a Neutron Flux-Trap intended for 5MW Tehran Research Reactor was designed. Fuel conversion from HEU to LEU in research reactors usually deprives the core from the high neutron flux. Therefore one has to look for a remedy in such situations otherwise radioisotope production, especially for those neutron demanding ones, falls down dramatically. The initiations of Neutron Flux-Trap idea comes true to face up this problem and provide an appropriate place inside or outside the core with sufficient neutron flux higher than the normal average level. To implement such a design, a number of codes and calculational tools have been used. At preliminary stage, WIMSD and EXTERMINATOR-II were used and then at later stages CITATION and MCNP codes were used for final design. Furthermore, SAND-II and ORIGEN were also employed for spectral analysis and radioisotope production calculations. Good agreements resulted with experiments and especially the case that Neutron Flux-Trap filled with ordinary water

  7. Computer aided design of fast neutron therapy units

    International Nuclear Information System (INIS)

    Gileadi, A.E.; Gomberg, H.J.; Lampe, I.

    1980-01-01

    Conceptual design of a radiation-therapy unit using fusion neutrons is presently being considered by KMS Fusion, Inc. As part of this effort, a powerful and versatile computer code, TBEAM, has been developed which enables the user to determine physical characteristics of the fast neutron beam generated in the facility under consideration, using certain given design parameters of the facility as inputs. TBEAM uses the method of statistical sampling (Monte Carlo) to solve the space, time and energy dependent neutron transport equation relating to the conceptual design described by the user-supplied input parameters. The code traces the individual source neutrons as they propagate throughout the shield-collimator structure of the unit, and it keeps track of each interaction by type, position and energy. In its present version, TBEAM is applicable to homogeneous and laminated shields of spherical geometry, to collimator apertures of conical shape, and to neutrons emitted by point sources or such plate sources as are used in neutron generators of various types. TBEAM-generated results comparing the performance of point or plate sources in otherwise identical shield-collimator configurations are presented in numerical form. (H.K.)

  8. Neutronics design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

    1983-01-01

    A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

  9. A Polyethylene Moderator Design for Auxiliary Ex-core Neutron Detector

    International Nuclear Information System (INIS)

    Lee, Hwan Soo; Shin, Ho Cheol; Bae, Seong Man

    2012-01-01

    The moderator of detector assembly in ENFMS (Excore Neutron Flux Monitoring System) plays a key role for slowing down from fast neutron to thermal neutron at outside of reactor vessel. Since neutron monitoring detector such as BF3, fission chamber detectors mostly responds to thermal neutron, moderator should be included to neutron detector assembly to detect more efficiently. Generally, resin has been used for moderator of detector in ENFMS of OPR1000 and APR1400, because resin has stable thermal resistance, availability and high neutron moderation characteristics due to the light atomic materials. In case of an auxiliary ex-core neutron detector, the polyethylene is suggested that polyethylene has a better moderator rather than resin, then, the amounts of moderator are reduced. This is important thing for auxiliary ex-core detector equipment at reactor, because the auxiliary equipment should affect minimally to another system. In this study, polyethylene moderator is designed for auxiliary ex-core neutron detector. To find out the optimal thickness of polyethylene moderator, preliminary simulation and experiments are performed. And sensitivity simulation for detector moderator at actual reactor is performed by DORT code

  10. Kartini Research Reactor prospective studies for neutron scattering application

    International Nuclear Information System (INIS)

    Widarto

    1999-01-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  11. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  12. Criteria design of the CAREM 25 reactor's core: neutronic aspects

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    The criteria that guided the design, from the neutronic point of view, of the CAREM reactor's core were presented. The minimum set of objectives and general criteria which permitted the design of the particular systems constituting the CAREM 25 reactor's core is detailed and stated. (Author) [es

  13. The Advanced Neutron Source design: A status report

    International Nuclear Information System (INIS)

    West, C.D.

    1992-01-01

    The Advanced Nuetron Source (ANS) facility is being designed as a user laboratory for all types of neutron-based research, centered around a nuclear fission reactor (D 2 O cooled, moderated, and reflected), operating at approximately 300 MW th . Safety, and especially passive safety features, have been emphasized throughout the design process

  14. BNL feasibility studies of spallation neutron sources

    International Nuclear Information System (INIS)

    Lee, Y.Y.; Ruggiero, A.G.; Van Steenbergen, A.; Weng, W.T.

    1995-01-01

    This paper is the summary of conceptual design studies of a 5 MW Pulsed Spallation Neutron Source (PSNS) conducted by an interdepartmental study group at Brookhaven National Laboratory. The study was made of two periods. First, a scenario based on the use of a 600 MeV Linac followed by two fast-cycling 3.6 GeV Synchrotrons was investigated. Then, in a subsequent period, the attention of the study was directed toward an Accumulator scenario with two options: (1) a 1.25 GeV normal conducting Linac followed by two Accumulator Rings, and (2) a 2.4 GeV superconducting Linac followed by a single Accumulator Ring. The study did not make any reference to a specific site

  15. Preclinical studies on gadolinium neutron capture therapy

    International Nuclear Information System (INIS)

    Akine, Yasuyuki

    1994-01-01

    Gadolinium neutron capture therapy is based on radiations (photons and electrons) produced in the tumor by a nuclear reaction between gadolinium and lower-energy neutrons. Studies with Chinese hamster cells have shown that the radiation effect resulting from gadolinium neutron capture reactions is mostly of low LET and that released electrons are the significant component in the over-all dose. Biological dosimetry revealed that the dose does not seem to increase in proportion to the gadolinium concentration, leading to a conclusion that there is a range of gadolinium concentrations most efficient for gadolinium neutron capture therapy. The in vivo studies with transplantable tumors in mice and rabbits have revealed that close contact between gadolinium and the cell is not necessarily required for cell inactivation and that gadolinium delivery selective to tumors is crucial. The results show that the potential of gadolinium neutron capture therapy as a therapeutic modality appears very promising. (author)

  16. Design of a portable directional neutron source finder

    International Nuclear Information System (INIS)

    Yamanishi, Hirokuni

    2005-01-01

    An instrument that determines the direction of a remote existing neutron source has been designed. This instrument combines a polyethylene block and four 3 He counter tubes. The advantages of the instrument are portability and good angular resolution. The count from the detector was varied with the neutron incident angle due to the moderator. Using this characteristic, the direction of the neutron source can be measured precisely by revising the axis of the instrument so that the difference between the four detectors measurements is minimized. Consequently, the direction of the central axis of the instrument in which the response difference of the four detectors reaches a minimum indicates the direction of the neutron source. The practical use of the instrument was demonstrated by 252 Cf source irradiation experiment and MCNP simulation

  17. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    International Nuclear Information System (INIS)

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 10 8 n/cm 2 · s. The fast neutron and gamma radiation KERMA factors are 10 x 10 -11 cGy·cm 2 /n epi and 20 x 10 -11 cGy·cm 2 /n epi , respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power

  18. Livermore intense neutron source: design concepts

    International Nuclear Information System (INIS)

    Davis, J.C.; Anderson, J.D.; Booth, R.; Logan, C.M.; Osher, J.E.

    1975-07-01

    The Lawrence Livermore laboratory proposes to build an irradiation facility containing several 14 MeV T(d,n) neutron sources for materials damage experimentation. A source strength of 4 x 10 13 n/s can be produced with 400 keV D + beam on the tritium in titanium target system now used on the Livermore Rotating Target Neutron Source (RTNS). To produce the desired source strength an accelerator which can deliver 150 mA of 400 keV D + ions must be built. For the target to remain within the time-temperature regime of the present system it must have a diameter of 46 cm and rotate at 5000 rpm. With a beam spot 1 cm fwhm the useful target lifetime is expected to be the 100 hours typical of the present system. A maximum flux of 1.5 x 10 13 n/cm 2 s will be attainable over a sample 1 mm thick by 8 mm in diameter. (U.S.)

  19. On the design of a cold neutron irradiator (CNI) for quantitative materials characterization

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Alexander Grover [Cornell Univ., Ithaca, NY (United States)

    1997-08-01

    A design study of a cold neutron irradiator (CNI) for materials characterization using prompt gamma-ray neutron activation analysis (PGNAA) is presented. Using 252Cf neutron sources in a block of moderator, a portion of which is maintained at a cryogenic temperature, the CNI employs cold neutrons instead of thermal neutrons to enhance the neutron capture reaction rate in a sample. Capture gamma rays are detected in an HPGe photon detector. Optimization of the CNI with respect to elemental sensitivity (counts per mg) is the primary goal of this design study. Monte Carlo simulation of radiation transport, by means of the MCNP code and the ENDF/B cross-section libraries, is used to model the CNI. A combination of solid methane at 22 K, room-temperature polyethylene, and room-temperature beryllium has been chosen for the neutron delivery subsystem of the CNI. Using four 250-microgram 252Cf neutron sources, with a total neutron emission rate of 2.3 x 109 neutrons/s, a thermal-equivalent neutron flux of 1.7 x 107 neutrons/cm2-s in an internally located cylindrical sample space of diameter 6.5 cm and height 6.0 cm is predicted by MCNP calculations. A cylindrical port with an integral annular collimator composed of bismuth, lead, polyethylene, and lithium carbonate, is located between the sample and the detector. Calculations have been performed of gamma-ray and neutron transport in the port and integral collimator with the objective of optimizing the statistical precision with which one can measure elemental masses in the sample while also limiting the fast neutron flux incident upon the HPGe detector to a reasonable level. The statistical precision with which one can measure elemental masses can be enhanced by a factor of between 2.3 and 5.3 (depending on the origin of the background gamma rays) compared with a neutron irradiator identical to the CNI except for the replacement of the cryogenic solid methane by room

  20. Conceptual design of a neutron camera for MAST Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Weiszflog, M., E-mail: matthias.weiszflog@physics.uu.se; Sangaroon, S.; Cecconello, M.; Conroy, S.; Ericsson, G.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, EURATOM-VR Association, Uppsala (Sweden); Keeling, D.; Martin, R. [CCFE, Culham Science Centre, Abingdon (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrae 2, D-85748 Garching (Germany)

    2014-11-15

    This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spherical Tokamak (MAST) Upgrade. The first one consists of two horizontal cameras, one equatorial and one vertically down-shifted by 65 cm. The second design, viewing the plasma in a poloidal section, also consists of two cameras, one radial and the other one with a diagonal view. Design parameters for the different cameras were selected on the basis of neutron transport calculations and on a set of target measurement requirements taking into account the predicted neutron emissivities in the different MAST Upgrade operating scenarios. Based on a comparison of the cameras’ profile resolving power, the horizontal cameras are suggested as the best option.

  1. Reactivity studies on the advanced neutron source

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Redmond, E.L. II; Fletcher, C.D.

    1990-01-01

    An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 x 10 19 m -2 s -1 is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria. 8 refs., 3 tabs

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

    Directory of Open Access Journals (Sweden)

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

    2016-10-01

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

  3. Design and Rationale for an In Situ Cryogenic Deformation Capability at a Neutron Source

    International Nuclear Information System (INIS)

    Livescu, V.; Clausen, B.; Sisneros, T.; Bourke, M.A.M.; Woodruff, T.R.; Vaidyanathan, R.; Notardonato, W.U.

    2004-01-01

    When performed in conjunction with neutron diffraction, in situ loading offers unique insights on microstructural deformation mechanisms. This is by virtue of the penetration and phase sensitivity of neutrons. At Los Alamos National Laboratory room and high temperature (up to 1500 deg. C) polycrystalline constitutive response is modeled using finite element and self-consistent models. The models are compared to neutron diffraction measurements. In doing so the implications of slip and creep to microstructural response have been explored. Recently we have been considering low temperature phenomena. This includes changes in deformation mechanisms such as the increased predilection for twinning over slip. Since this is associated with measurable texture changes as well as microstructural strain effects, it is well suited for study using neutron diffraction. This paper outlines the design and rationale for a cryogenic loading capability that will be used on the Spectrometer for MAterials Research at Temperature and Stress (SMARTS) at the Los Alamos Neutron Science Center (LANSCE)

  4. Design of a boron neutron capture enhanced fast neutron therapy assembly

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhonglu [Georgia Inst. of Technology, Atlanta, GA (United States)

    2006-12-01

    The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm2 treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm2 collimation was 21.9% per 100-ppm 10B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm2 fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm2 collimator. Five 1.0-cm thick 20x20 cm2 tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm 10B) to measure dose due to boron neutron capture. The

  5. Design of 6 Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)

    2012-01-15

    The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.

  6. Neutron shielding and its impact on the ITER machine design

    International Nuclear Information System (INIS)

    Daenner, W.; El Guebaly, L.; Sawan, M.; Gohar, Y.; Maki, K.; Rado, V.; Schchipakin, O.; Zimin, S.

    1991-01-01

    This paper describes the efforts made in the frame of the ITER project to analyze the shielding of the superconducting magnets. First, the radiation limits to be achieved are specified as well as the neutron source in terms of wall loading on the first wall of the machine. Then the general shield concept is explained, including the most essential details of the various shield components. A brief section is devoted to the calculational tools, the data base, and the safety factors to be applied to the results obtained. The neutronics models of four different configurations are summarized as they were used to study the most critical parts of the machine. This section is followed by a presentation of the most important results from one-, two- and three-dimensional calculations. They are given for both the reference design and an improved one in which the critical regions are reinforced with respect to their shielding capability. It is concluded that the ITER shield layout just marginally meets the stated limits provided that some tungsten is included in the critical regions. A slight revision of the overall machine dimensions with the aim to achieve a less complex shield and a higher margin with respect to the limits is, however, seen the better solution. (orig.)

  7. Neutronic studies of a 233U breeder

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1978-09-01

    Neutronic calculations have been carried out to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (>1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The performance of these two blankets is discussed in terms of their energy multiplication, tritium breeding and fissile fuel production. The neutronic calculations have been done for two neutron libraries, the ENDF/B-IV and the ENDL with differences no larger than 10% in the results. An estimate is given of the number of equivalent thermal power fission reactors (LWR, HWR, SSCR, and HTGR) that these fusion breeders can fuel

  8. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  9. Study of neutron spectrometers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kaellne, Jan

    2005-11-15

    A review is presented of the developments in the field of neutron emission spectrometry (NES) which is of relevance for identifying the role of NES diagnostics on ITER and selecting suitable instrumentation. Neutron spectrometers will be part of the ITER neutron diagnostic complement and this study makes a special effort to examine which performance characteristics the spectrometers should possess to provide the best burning plasma diagnostic information together with neutron cameras and neutron yield monitors. The performance of NES diagnostics is coupled to how much interface space can be provided which has lead to an interest to find compact instruments and their NES capabilities. This study assesses all known spectrometer types of potential interest for ITER and makes a ranking of their performance (as demonstrated or projected), which, in turn, are compared with ITER measurement requirements as a reference; the ratio of diagnostic performance to interface cost for different spectrometers is also discussed for different spectrometer types. The overall result of the study is an assessment of which diagnostic functions neutron measurements can provide in burning plasma fusion experiments on ITER and the role that NES can play depending on the category of instrument installed. Of special note is the result that much higher quality diagnostic information can be obtained from neutron measurements with total yield monitors, profile flux cameras and spectrometers when the synergy in the data is considered in the analysis and interpretation.

  10. Neutronic calculations in support of the design of the ITER High Resolution Neutron Spectrometer

    International Nuclear Information System (INIS)

    Moro, F.; Esposito, B.; Marocco, D.; Villari, R.; Petrizzi, L.; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Dapena, M.

    2011-01-01

    This paper presents the results of neutronic calculations performed to address important issues related to the optimization of the ITER HRNS (High resolution Neutron Spectrometer) design, in particular concerning the definition of the collimator and the choice of the detector system. The calculations have been carried out using the MCNP5 Monte Carlo code in a full 3-D geometry. The HRNS collimation system has been included in the latest MCNP ITER 40 o model (Alite-4). The ITER scenario 2 reference DT plasma fusion neutron source peaked at 14.1 MeV with Gaussian energy distribution has been used. Neutron fluxes and energy spectra (>1 MeV) have been evaluated at different positions along the HRNS collimator and at the detector location. The noise-to-signal ratio (i.e. the ratio of collided to uncollided neutrons), the breakdown of the collided spectrum into its components, the dependency on the first wall aperture and the gamma-ray spectra at the detector position have also been analyzed. The impact of the results on the design of the HRNS diagnostic system is discussed.

  11. 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.

  12. Software for simulation and design of neutron scattering instrumentation

    DEFF Research Database (Denmark)

    Bertelsen, Mads

    designed using the software. The Union components uses a new approach to simulation of samples in McStas. The properties of a sample are split into geometrical and material, simplifying user input, and allowing the construction of complicated geometries such as sample environments. Multiple scattering...... from conventional choices. Simulation of neutron scattering instrumentation is used when designing instrumentation, but also to understand instrumental effects on the measured scattering data. The Monte Carlo ray-tracing package McStas is among the most popular, capable of simulating the path of each...... neutron through the instrument using an easy to learn language. The subject of the defended thesis is contributions to the McStas language in the form of the software package guide_bot and the Union components.The guide_bot package simplifies the process of optimizing neutron guides by writing the Mc...

  13. Basic Design Report of DC-TOF Inelastic Neutron Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    So, Ji Yong; Park, Je Geun; Moon, Myung Kook; Cho, Sang Jin; Choi, Yung Hyun; Lee, Chang Hee

    2006-04-15

    We made Basic designs of neutron guide, choppers, and detectors in order to optimize the design parameters of DC-TOF to be built in the HANARO Cold Neutron Guide Hall. In addition, we calculated the expected performance of DC-TOF using Monte Carlo simulations and evaluated the properties of neutron beam. Based on the results we obtained, we have compared the expected performance of the DC-TOF with those of existing instruments overseas. In conclusion, we believe that we will be able to construct the DC-TOF at HANARO as one of the best instruments of its kinds and it will become an invaluable instrument to researchers in the related field.

  14. Design considerations for primary neutron beam collimation on the Spallation Neutron Source

    International Nuclear Information System (INIS)

    Howells, W.S.

    1980-09-01

    A scheme for the design of primary neutron beam collimation is presented which is based on ray diagrams. The practical application of the ideas is outlined and the influence of various constraints such as beam shutters is discussed. The ideas are illustrated with examples which include the layouts for some typical instruments. (author)

  15. Study on condensed media with polarized neutrons

    International Nuclear Information System (INIS)

    Drabkin, G.M.

    1974-01-01

    In this paper are considered the results of a study of the secondary magnetic superstructure of ferromagnets in the phase transition region by means of polarized neutrons. The results obtained are compared with experimental data

  16. Neutron scattering studies of solid electrolytes

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1976-01-01

    The role which neutron scattering can play in determining the nature of the disorder and the conducting mechanism in the solid electrolytes is discussed. First, some of the general formalism for elastic and inelastic neutron scattering is reviewed, and the quantities which can be measured are pointed out. Then the application of neutron scattering to the studies of three different problems is examined; the anion disorder in the fluorite system, the dynamical behavior in beta-alumina, and the cation diffusion in αAgI are discussed. 8 figures

  17. Neutron radiation damage studies on silicon detectors

    International Nuclear Information System (INIS)

    Li, Zheng; Chen, W.; Kraner, H.W.

    1990-10-01

    Effects of neutron radiation on electrical properties of Si detectors have been studied. At high neutron fluence (Φ n ≥ 10 12 n/cm 2 ), C-V characteristics of detectors with high resistivities (ρ ≥ 1 kΩ-cm) become frequency dependent. A two-trap level model describing this frequency dependent effect is proposed. Room temperature anneal of neutron damaged (at LN 2 temperature) detectors shows three anneal stages, while only two anneal stages were observed in elevated temperature anneal. 19 refs., 14 figs

  18. Study of neutron fields around an intense neutron generator.

    Science.gov (United States)

    Kicka, L; Machrafi, R; Miller, A

    2017-12-01

    Neutron fields in the vicinity of the newly built neutron facility, at the University of Ontario Institute of Technology (UOIT), have been investigated in a series of Monte Carlo simulations and measurements. The facility hosts a P-385 neutron generator based on a deuterium-deuterium fusion reaction. The neutron fluence at different locations around the neutron generator facility has been simulated using MCNPX 2.7E Monte Carlo particle transport program. To characterize neutron fields, three neutron sources were modeled with distributions corresponding to different incident deuteron energies of 90kV, 110kV, and 130kV. Measurements have been carried out to determine the dose rate at locations adjacent to the generator using bubble detectors (BDs). The neutron intensity was evaluated and the total dose rates corresponding to different applied acceleration potentials were estimated at various locations. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Studies in precise neutron optics at JAERI

    International Nuclear Information System (INIS)

    Tomimitsu, Hiroshi

    1994-01-01

    Studies in the field of so-called 'Precise Neutron Optics' at JAERI were reported, since the beginning of 1970. It started with the photographic detection of the 'Pendellusung Fringes' in a wedge-shaped Si crystal, promoted by Prof.K.Kohra. Neutron diffraction topography was also tried and used for the 'direct' observation of the substructure in a Cu-5%Ge single crystal, carried out by the present author. So-called 'VSANS' (Very Small Angle Neutron Scattering) was also tried, with the angular resolution better than 0.1 sec. of arc, for the structural observation of several kinds of 'amorphous' materials, such as neutron-irradiated silica glasses, Pb-containing glasses and several ribbon-shaped amorphous alloys, carried out by Dr.K.Doi. Recently with the co-operation with Prof.S.Kikuta's group, some interferometric experiments were tried, such as the detection of the double-refraction phenomenon of the neutron in the magnetic materials, carried out by Dr.S.Nakatani. The apparatus for precise neutron optics and topography (PNO) was constructed at the JRR-3M in JAERI in 1992, which was just properly made for the studies in the present thema. With the PNO, the characterization of Ni-Ti multilayer mirror for the interferometer with the cold/ ultra-cold neutrons was carried out, the results being reported in the other article of this symposium by Dr.H.Funahashi. Through the test use of the PNO with the neutron interferometry, carried out by Dr.Y.Hasegawa, it was revealed that the PNO is the best facility in the world for the neutron interferometry; with the datas of the count ratio upto 20 cps and the contrast more than 40%. Dr.K.Aizawa is now refining the VSANS-technique on the PNO, for the precise observation of the precipitation behaviour of high density small particles such as metatic alloy-system. (author)

  20. Neutron streaming analysis for shield design of FMIT Facility

    International Nuclear Information System (INIS)

    Carter, L.L.

    1980-12-01

    Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

  1. The Neutron star Interior Composition Explorer (NICER): design and development

    DEFF Research Database (Denmark)

    Gendreau, Keith C.; Arzoumanian, Zaven; Adkins, Phillip W.

    2016-01-01

    During 2014 and 2015, NASA's Neutron star Interior Composition Explorer (NICER) mission proceeded successfully through Phase C, Design and Development. An X-ray (0.2-12 keV) astrophysics payload destined for the International Space Station, NICER is manifested for launch in early 2017...

  2. Design of the 50 kW neutron converter for SPIRAL2 facility

    Energy Technology Data Exchange (ETDEWEB)

    Avilov, M.S. [Budker Institute of Nuclear Physics, 630090 Novosibirsk, SB RAS (Russian Federation); Tecchio, L.B., E-mail: tecchio@lnl.infn.i [Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Titov, A.T. [Boreskov Institute of Catalysis, 630090 Novosibirsk, SB RAS (Russian Federation); Tsybulya, V.S. [Trofimuk Institute of Geology, 630090 Novosibirsk, SB RAS (Russian Federation); Zhmurikov, E.I. [Budker Institute of Nuclear Physics, 630090 Novosibirsk, SB RAS (Russian Federation)

    2010-06-21

    SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

  3. Design of the 50 kW neutron converter for SPIRAL2 facility

    International Nuclear Information System (INIS)

    Avilov, M.S.; Tecchio, L.B.; Titov, A.T.; Tsybulya, V.S.; Zhmurikov, E.I.

    2010-01-01

    SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

  4. Introduction to Neutron Coincidence Counter Design Based on Boron-10

    Energy Technology Data Exchange (ETDEWEB)

    Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Siciliano, Edward R.

    2012-01-22

    The Department of Energy Office of Nonproliferation Policy (NA-241) is supporting the project 'Coincidence Counting With Boron-Based Alternative Neutron Detection Technology' at Pacific Northwest National Laboratory (PNNL) for development of an alternative neutron coincidence counter. The goal of this project is ultimately to design, build and demonstrate a boron-lined proportional tube based alternative system in the configuration of a coincidence counter. This report, providing background information for this project, is the deliverable under Task 1 of the project.

  5. Study of a transportable neutron radiography system

    International Nuclear Information System (INIS)

    Souza, S.N.A. de.

    1991-05-01

    This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs

  6. Study of neutron-deficient Sn isotopes

    International Nuclear Information System (INIS)

    Auger, G.

    1982-05-01

    The formation of neutron deficient nuclei by heavy ion reactions is investigated. The experimental technique is presented, and the results obtained concerning Sn et In isotopes reported: first excited states of 106 Sn, high spin states in 107 Sn and 107 In; Yrast levels of 106 Sn, 107 Sn, 108 Sn; study of neutron deficient Sn and In isotopes formed by the desintegration of the compound nucleus 112 Xe. All these results are discussed [fr

  7. Neutron Scattering studies of magnetic molecular magnets

    International Nuclear Information System (INIS)

    Chaboussant, G.

    2009-01-01

    This work deals with inelastic neutron scattering studies of magnetic molecular magnets and focuses on their magnetic properties at low temperature and low energies. Several molecular magnets (Mn 12 , V 15 , Ni 12 , Mn 4 , etc.) are reviewed. Inelastic neutron scattering is shown to be a perfectly suited spectroscopy tool to -a) probe magnetic energy levels in such systems and -b) provide key information to understand the quantum tunnel effect of the magnetization in molecular spin clusters. (author)

  8. Development of nuclear design criteria for neutron spallation sources

    Energy Technology Data Exchange (ETDEWEB)

    Sordo, F.; Abanades, A. [E.T.S. Industriales, Madrid Polytechnic University, UPM, J.Gutierrez Abascal, 2 -28006 Madrid (Spain)

    2008-07-01

    Spallation neutron sources allow obtaining high neutronic flux for many scientific and industrial applications. In recent years, several proposals have been made about its use, notably the European Spallation Source (ESS), the Japanese Spallation Source (JSNS) and the projects of Accelerator-Driven Subcritical reactors (ADS), particularly in the framework of EURATOM programs. Given their interest, it seems necessary to establish adequate design basis for guiding the engineering analysis and construction projects of this kind of installations. In this sense, all works done so far seek to obtain particular solutions to a particular design, but there has not been any general development to set up an engineering methodology in this field. In the integral design of a spallation source, all relevant physical processes that may influence its behaviour must be taken into account. Neutronic aspects (emitted neutrons and their spectrum, generation performance..), thermomechanical (energy deposition, cooling conditions, stress distribution..), radiological (spallation waste activity, activation reactions and residual heat) and material properties alteration due to irradiation (atomic displacements and gas generation) must all be considered. After analysing in a systematic manner the different options available in scientific literature, the main objective of this thesis was established as making a significant contribution to determine the limiting factors of the main aspects of spallation sources, its application range and the criteria for choosing optimal materials. To achieve this goal, a series of general simulations have been completed, covering all the relevant physical processes in the neutronic and thermal-mechanical field. Finally, the obtained criteria have been applied to the particular case of the design of the spallation source of subcritical reactors PDX-ADS and XT-ADS. These two designs, developed under the European R and D Framework Program, represent nowadays

  9. Development of nuclear design criteria for neutron spallation sources

    International Nuclear Information System (INIS)

    Sordo, F.; Abanades, A.

    2008-01-01

    Spallation neutron sources allow obtaining high neutronic flux for many scientific and industrial applications. In recent years, several proposals have been made about its use, notably the European Spallation Source (ESS), the Japanese Spallation Source (JSNS) and the projects of Accelerator-Driven Subcritical reactors (ADS), particularly in the framework of EURATOM programs. Given their interest, it seems necessary to establish adequate design basis for guiding the engineering analysis and construction projects of this kind of installations. In this sense, all works done so far seek to obtain particular solutions to a particular design, but there has not been any general development to set up an engineering methodology in this field. In the integral design of a spallation source, all relevant physical processes that may influence its behaviour must be taken into account. Neutronic aspects (emitted neutrons and their spectrum, generation performance..), thermomechanical (energy deposition, cooling conditions, stress distribution..), radiological (spallation waste activity, activation reactions and residual heat) and material properties alteration due to irradiation (atomic displacements and gas generation) must all be considered. After analysing in a systematic manner the different options available in scientific literature, the main objective of this thesis was established as making a significant contribution to determine the limiting factors of the main aspects of spallation sources, its application range and the criteria for choosing optimal materials. To achieve this goal, a series of general simulations have been completed, covering all the relevant physical processes in the neutronic and thermal-mechanical field. Finally, the obtained criteria have been applied to the particular case of the design of the spallation source of subcritical reactors PDX-ADS and XT-ADS. These two designs, developed under the European R and D Framework Program, represent nowadays

  10. Development of a neutron converter for studies of neutron-induced fission fragments at the IGISOL facility

    CERN Document Server

    Lantz, M; Al-Adili, A; Jokinen, A; Kolhinen, V; Mattera, A; Rinta-Antila, S; Penttilä, H; Pomp, S; Rakoupoulos, V; Simutkin, V; Solders, A

    2014-01-01

    The ERINDA funded scientific visit has enabled the groups at U ppsala Uni- versity and University of Jyväskylä to work closer together on the design of a neutron converter that will be used as neutron source in fissi on yield studies at the IGISOL-JYFLTRAP facility at the University of Jyväsk ylä. The design is based on simulations with both deterministic codes and Mo nte Carlo codes, and an ERINDA funded benchmark measurement. In order to obta in a com- petitive count rate the fission targets will be placed very cl ose to the neutron converter. The intention is to have a flexible design that wil l enable neutron fields with different energy distributions. In this report t he progression and the present status of the design work will be discussed, togethe r with an outlook of the future plans

  11. The design of the cold neutron source of the OPAL reactor

    International Nuclear Information System (INIS)

    Rechiman, L.M.; Bonetto, Fabian J.; Buscaglia, Gustavo C.

    2007-01-01

    The present work describes the conceptual design process of the first cold neutron source developed by INVAP for the nuclear research reactor OPAL. The analysis begins from the requirements given by the client and continues with the chosen solutions. Furthermore, we studied how impact in the design the fully illuminated constraint with the finite remote source model. (author) [es

  12. Design, construction, and calibration of a nonfocusing neutron spectrometer

    International Nuclear Information System (INIS)

    Storey, W.

    1974-12-01

    A fourteen-channel time-resolved neutron spectrometer with associated Faraday cup has been designed and constructed for use in the field. A neutron energy range of 9.5 to 15 MeV is covered. Both instruments detect protons elastically scattered from a thin hydrogenous foil in interaction with the neutron beam, with magnetic analysis of the protons by the spectrometer. The design requirements of small size and weight and 0.6 to 0.7 MeV resolution have been met. Following a description of the instrument and of its geometry, there is a detailed presentation of the design and construction of the instrument. The section on instrument performance is concerned with the comparison between predicted performance based upon computation, in which the magnet is of primary interest, and upon measured performance based upon a calibration experiment, which is given a general description in Appendix A. Software used mainly for signal prediction and unfolding, for both the neutron spectrometer and Faraday cup, is described

  13. Design of active-neutron fuel rod scanner

    International Nuclear Information System (INIS)

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

    1996-01-01

    An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

  14. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  15. Study of an individual neutron dosimeter

    International Nuclear Information System (INIS)

    Debeauvais, M.; Tripier, J.

    1976-01-01

    A dosimeter using Kodak LR 115 cellulose nitrate as detecting material was designed. It serves to determine 3 neutron energy ranges. The 6 Li(n,α)t reaction is used for the thermal region, the sensitivity being 0.2mrads to 1 rad for neutron energies between thermal and 0.05eV. The same reaction defines the 0.05eV to 1000eV energy range but the detection system is placed inside a cadmium screen; the sensitivity is 0.2 to 500rads. Finally above 1MeV the neutron reactions used are those on the detector components themselves, i.e. elastic collisions and (nα) reactions on carbon, nitrogen and oxygen nuclei. Detection is possible between 0.7 and 700 rads [fr

  16. Neutronic feasibility design of a small long-life HTR

    International Nuclear Information System (INIS)

    Ding Ming; Kloosterman, Jan Leen

    2011-01-01

    Highlights: ► We propose the neutronic feasibility design of a small, long lifetime and transportable HTR. ► Comparison of cylindrical, annular and scatter cores of the small block-type HTR. ► The design of the scatter core effectively reduces the number of the fuel block and increases the lifetime and burnup of the reactor. - Abstract: Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U

  17. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  18. Optimum design of exploding pusher target to produce maximum neutrons

    International Nuclear Information System (INIS)

    Kitagawa, Y.; Miyanaga, N.; Kato, Y.; Nakatsuka, M.; Nishiguchi, A.; Yabe, T.; Yamanaka, C.

    1985-03-01

    Exploding pusher target experiments have been conducted with the 1.052-μm GEKKO MII two-beam glass laser system to design an optimum target, which couples to the incident laser light most effectively to produce the maximum neutrons. Since hot electrons preheat the shell entirely in spite of strongly nonuniform irradiation, a simple model can design the optimum target, of which the shell/fuel interface is accelerated to 0.5 to 0.7 times the initial radius within a laser pulse. A 2-dimensional computer simulation supports this target design. The scaling of the neutron yield N with the laser power P is N ∝ P 2.4±0.4 . (author)

  19. Magnetic structures: neutron diffraction studies

    International Nuclear Information System (INIS)

    Bouree-Vigneron, F.

    1990-01-01

    Neutron diffraction is often an unequivocal method for determining magnetic structures. Here we present some typical examples, stressing the sequence through experiments, data analysis, interpretation and modelisation. Two series of compounds are chosen: Tb Ni 2 Ge 2 and RBe 13 (R = Gd, Tb, Dy, Ho, Er). Depending on the nature of the elements, the magnetic structures produced can be commensurate, incommensurate or even show a transition between two such phases as a function of temperature. A model, taking magnetic exchange and anisotropy into account, will be presented in the case of commensurate-incommensurate magnetic transitions in RBe 13

  20. Study on neutron activation analysis

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Cho, Seung Yeon

    1993-01-01

    Environmental samples were analyzed quantitatively by neutron activation analysis using high resolution γ-ray spectrometry. The accuracy and precision of the method were checked by the analysis of reference materials, Urban Particulate Matter (NBS SRM 1648) and Coalfly ash (NBS SRM 1633a). Airborne particulates collected for 6 months with low volume air sampler at the outer area of Seoul were analyzed as the start of full scale airborne particulates research. We analyzed 19 trace elements from the samples and the NAA method was confirmed to be utilized for environmental pollution research. (Author)

  1. Optimizing a three-element core design for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

    Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

  2. A standardized method for beam design in neutron capture therapy

    International Nuclear Information System (INIS)

    Storr, G.J.: Harrington, B.V.

    1993-01-01

    A desirable end point for a given beam design for Neutron Capture Therapy (NCT) should be quantitative description of tumour control probability and normal tissue damage. Achieving this goal will ultimately rely on data from NCT human clinical trials. Traditional descriptions of beam designs have used a variety of assessment methods to quantify proposed or installed beam designs. These methods include measurement and calculation of open-quotes free fieldclose quotes parameters, such as neutron and gamma flux intensities and energy spectra, and figures-of-merit in tissue equivalent phantoms. The authors propose here a standardized method for beam design in NCT. This method would allow all proposed and existing NCT beam facilities to be compared equally. The traditional approach to determining a quantitative description of tumour control probability and normal tissue damage in NCT research may be described by the following path: Beam design → dosimetry → macroscopic effects → microscopic effects. Methods exist that allow neutron and gamma fluxes and energy dependence to be calculated and measured to good accuracy. By using this information and intermediate dosimetric quantities such as kerma factors for neutrons and gammas, macroscopic effect (absorbed dose) in geometries of tissue or tissue-equivalent materials can be calculated. After this stage, for NCT the data begins to become more sparse and in some areas ambiguous. Uncertainties in the Relative Biological Effectiveness (RBE) of some NCT dose components means that beam designs based on assumptions considered valid a few years ago may have to be reassessed. A standard method is therefore useful for comparing different NCT facilities

  3. A Study on the Thermal Neutron Filter for the Irradiation of Electronic Materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Sung Ryul; Park, Seung Jae; Shin, Yoon Taeg; Cho, Man Soon; Cho, Kee Nam [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The representative example is a technique of making the semiconductor with the transmutation using the pure Si. This NTD (Neutron Transmutation Doping) Si is used as a high-quality semiconductor because it has a uniform resistance. Likewise, the electronic materials are being investigated to improve the performance of material using the neutron irradiation method. The mechanism for reaction between the electronic materials and the neutrons depends on the energy of the neutron. Capturing reaction by thermal neutrons causes the transmutation and a lot of defects are made by fast neutrons. The study for the effect by such neutron energy is necessary to understand the performance improvement of the irradiated electronic materials. The thermal neutron filter was investigated to be used for the irradiation of electronic materials at HANARO. IP irradiation hole was selected and the irradiation device was designed. The analysis was conducted considering four candidate materials.

  4. A neutron beam polarizer for study of parity violation in neutron-nucleus interactions

    International Nuclear Information System (INIS)

    Penttilae, S.I.; Bowman, J.D.; Frankle, C.M.; Seestrom, S.J.; Yen, Yi-Fen; Delheij, P.P.J.; Haase, D.G.; Postma, H.

    1994-01-01

    A dynamically-polarized proton target operating at 5 Tesla and 1 K has been built to, neutron beam for studies of parity violation in compound-nuclear resonances. Nearly 0.9 proton polarization was obtained in an electron-beam irradiated ammonia target. This was used to produce a neutron beam polarization of 0.7 at epithermal energies. The combination of the polarized proton target and the LANSCE spallation neutron source produces the most intense pulsed polarized epithermal neutron beam in the world. The neutron-beam polarizer is described and methods to determine neutron beam polarization are presented

  5. A neutron beam polarizer for study of parity violation in neutron-nucleus interactions

    International Nuclear Information System (INIS)

    Penttilae, S.I.; Bowman, J.D.; Delheij, P.P.; Frankle, C.M.; Haase, D.G.; Postma, H.; Seestrom, S.J.; Yen, Y.

    1995-01-01

    A dynamically-polarized proton target operating at 5 Tesla and 1 K has been built to polarize an epithermal neutron beam for studies of parity violation in compound-nuclear resonances. Nearly 0.9 proton polarization was obtained in an electron-beam irradiated ammonia target. This was used to produce a neutron beam polarization of 0.7 at epithermal energies. The combination of the polarized proton target and the LANSCE spallation neutron source produces the most intense pulsed polarized epithermal neutron beam in the world. The neutron-beam polarizer is described and methods to determine neutron beam polarization are presented. copyright 1995 American Institute of Physics

  6. 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...

  7. Design and tests of an adaptive focusing neutron guide

    International Nuclear Information System (INIS)

    Valicu, Roxana Georgiana

    2012-01-01

    This work contains the Monte Carlo Simulations, as well as the first tests with an adaptive focusing neutron guide for creating a focus that does not depend on the wavelength of the incoming neutrons. All known neutron guides consist of a rectangular shape, built out of four glass plates. The inner side of the guide is coated with a complex structure of metal layers. This reflects and guides the neutrons (in analogy with the reflection of the light). For beam focusing neutron guides with fixed curvature can be built. For most experiments it is important that the beam is focused on to a small surface of the sample. In the case of focusing guides with fixed curvature it has been observed that the focusing (dimension and position of the beam focus) is wavelength dependent. This is why for measurements that are performed with different wavelengths it is very important to change the curvature of the neutron guide in order to obtain optimal results. In this work we have designed, constructed and tested a guide where we can change the curvature during the experiment. In this way we can obtain a variable curvature in horizontal as well as in vertical direction. For a curvature in the horizontal or vertical direction it is not necessary to move all four walls, only two of the opposed plates. The element that changes the curvature of the guide consists of an acting element (piezomotor) as well as a rod that can be operated by the piezomotor and that acts through a lever onto the plate. The action of a force and a consecutive torsion momentum at the free end of the plate changes the curvature of the whole plate in an almost parabolic way. Making use of the Monte Carlo simulations we were able to determine the optimal curvature for each wavelength of a neutron guide for the spectrometer TOFTOF installed at the Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II). First tests have shown that with an adaptive focusing guide one can gain up to a factor three in intensity at

  8. Design and tests of an adaptive focusing neutron guide

    Energy Technology Data Exchange (ETDEWEB)

    Valicu, Roxana Georgiana

    2012-08-23

    This work contains the Monte Carlo Simulations, as well as the first tests with an adaptive focusing neutron guide for creating a focus that does not depend on the wavelength of the incoming neutrons. All known neutron guides consist of a rectangular shape, built out of four glass plates. The inner side of the guide is coated with a complex structure of metal layers. This reflects and guides the neutrons (in analogy with the reflection of the light). For beam focusing neutron guides with fixed curvature can be built. For most experiments it is important that the beam is focused on to a small surface of the sample. In the case of focusing guides with fixed curvature it has been observed that the focusing (dimension and position of the beam focus) is wavelength dependent. This is why for measurements that are performed with different wavelengths it is very important to change the curvature of the neutron guide in order to obtain optimal results. In this work we have designed, constructed and tested a guide where we can change the curvature during the experiment. In this way we can obtain a variable curvature in horizontal as well as in vertical direction. For a curvature in the horizontal or vertical direction it is not necessary to move all four walls, only two of the opposed plates. The element that changes the curvature of the guide consists of an acting element (piezomotor) as well as a rod that can be operated by the piezomotor and that acts through a lever onto the plate. The action of a force and a consecutive torsion momentum at the free end of the plate changes the curvature of the whole plate in an almost parabolic way. Making use of the Monte Carlo simulations we were able to determine the optimal curvature for each wavelength of a neutron guide for the spectrometer TOFTOF installed at the Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II). First tests have shown that with an adaptive focusing guide one can gain up to a factor three in intensity at

  9. Neutron-scattering studies of chromatin

    International Nuclear Information System (INIS)

    Bradbury, E.M.; Baldwin, J.P.; Carpenter, B.G.; Hjelm, R.P.; Hancock, R.; Ibel, K.

    1976-01-01

    It is clear that a knowledge of the basic molecular structure of chromatin is a prerequisite for any progress toward an understanding of chromosome organization. With a two-component system, protein and nucleic acid, neutrons have a particularly powerful application to studies of the spatial arrangements of these components because of the ability, by contrast matching with H 2 O-D 2 O mixtures, to obtain neutron-scattering data on the individual components. With this approach it has been shown that the neutron diffraction of chromatin is consistent with a ''beads on a string'' model in which the bead consists of a protein core with DNA coiled on the outside. However, because chromatin is a gel and gives limited structural data, confirmation of such a model requires extension of the neutron studies by deuteration of specific chromatin components and the isolation of chromatin subunits. Although these studies are not complete, the neutron results so far obtained support the subunit model described above

  10. Structural Analysis and Seismic Design for Cold Neutron Laboratory Building

    International Nuclear Information System (INIS)

    Wu, Sangik; Kim, Y. K.; Kim, H. R.

    2007-05-01

    This report describes all the major results of the dynamic structural analysis and seismic design for the Cold Neutron Laboratory Building which is classified in seismic class II. The results are summarized of the ground response spectrum as seismic input loads, mechanical properties of subsoil, the buoyancy stability due to ground water, the maximum displacement of the main frame under the seismic load and the member design. This report will be used as a basic design report to maintenance its structural integrity in future

  11. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Sagara, A.; Goto, T.; Yanagi, N.; Masuzaki, S.; Tamura, H.; Miyazawa, J.; Muroga, T., E-mail: teru@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: Neutronics design study has been performed in a newly started conceptual design activity for a helical type DEMO reactor FFHR-d1. Features of the FFHR-d1 design are enlargement of the basic configurations of reactor components and extrapolation of plasma parameters from those of the helical type plasma experimental machine Large Helical Device (LHD) to achieve the highest feasibility. From the neutronics point of view, a blanket space of FFHR-d1 is severely limited at the inboard of the torus. This is due to the core plasma position shifting to the inboard side under the confinement condition extrapolated from LHD. The first step of the neutronics investigation using the MCNP code has been performed with a simple torus model simulating thin inboard blanket space. A Flibe+Be/Ferritic steel breeding blanket showed preferable performances for both tritium breeding and shielding, and has been adapted as a reference blanket system for FFHR-d1. The investigations indicate that a combination of a 15 cm thick breeding blanket, 55 cm thick WC+B4C shield, i.e., the blanket space of 70 cm, could suppress the fast neutron flux and nuclear heating in the helical coils to the design targets for the neutron wall loading of 1.5 MW/m{sup 2}. Since the outboard side can provide a large space for a 60 cm thick breeding blanket, a fully-covered tritium breeding ratio (TBR) of 1.31 has been obtained in the simple torus model. The neutronics design study has proceeded to the second step using a 3-D helical reactor model. The most important issue in the 3-D neutronics design is a compatibility with the helical divertor design. To achieve a higher TBR and shielding performance, the core plasma has to be covered by the breeding blanket layers as possible. However, the dimensions of the blanket layers are limited by magnetic field lines connecting an edge of the core plasma and divertor pumping ports. After repeating modification of the blanket configuration, the global TBR of 1

  12. Neutron-Induced Charged Particle Studies at LANSCE

    Science.gov (United States)

    Lee, Hye Young; Haight, Robert C.

    2014-09-01

    Direct measurements on neutron-induced charged particle reactions are of interest for nuclear astrophysics and applied nuclear energy. LANSCE (Los Alamos Neutron Science Center) produces neutrons in energy of thermal to several hundreds MeV. There has been an effort at LANSCE to upgrade neutron-induced charged particle detection technique, which follows on (n,z) measurements made previously here and will have improved capabilities including larger solid angles, higher efficiency, and better signal to background ratios. For studying cross sections of low-energy neutron induced alpha reactions, Frisch-gridded ionization chamber is designed with segmented anodes for improving signal-to-noise ratio near reaction thresholds. Since double-differential cross sections on (n,p) and (n,a) reactions up to tens of MeV provide important information on deducing nuclear level density, the ionization chamber will be coupled with silicon strip detectors (DSSD) in order to stop energetic charged particles. In this paper, we will present the status of this development including the progress on detector design, calibrations and Monte Carlo simulations. This work is funded by the US Department of Energy - Los Alamos National Security, LLC under Contract DE-AC52-06NA25396.

  13. Neutron scattering studies on frustrated magnets

    International Nuclear Information System (INIS)

    Arima, Taka-hisa

    2013-01-01

    A lot of frustrated magnetic systems exhibit a nontrivial magnetic order, such as long-wavelength modulation, noncollinear, or noncoplanar order. The nontrivial order may pave the way for the novel magnetic function of matter. Neutron studies are necessary to determine the magnetic structures in the frustrated magnetic systems. In particular, spin-polarized neutron scattering is a useful technique for the investigation of the novel physical properties relevant to the nontrivial spin arrangement. Here some neutron studies on a multiferroic perovskite manganese oxide system are demonstrated as a typical case. The frustrated magnetic systems may also a playground of novel types of local magnetic excitations, which behave like particles in contrast to the magnetic waves. It is becoming a good challenge to study such particle-type magnetic excitations relevant to the magnetic frustration. (author)

  14. Neutron depolarization studies on magnetization process using pulsed polarized neutrons

    International Nuclear Information System (INIS)

    Mitsuda, Setsuo; Endoh, Yasuo

    1985-01-01

    Neutron depolarization experiments investigating the magnetization processes have been performed by using pulsed polarized neutrons for the first time. Results on both quenched and annealed ferromagnets of Fe 85 Cr 15 alloy indicate the significant difference in the wavelength dependence of depolarization between them. It also constitutes the experimental demonstration of the theoretical prediction of Halpern and Holstein. (author)

  15. Design of ex-vessel neutron monitor for ITER

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Yamauchi, Michinori; Kasai, Satoshi; Ebisawa, Katsuyuki; Walker, Chris

    2002-07-01

    A neutron flux monitor has been designed by using 235 U fission chambers to be installed outside the vacuum vessel of ITER. We investigated moderator materials to get flat energy response the responses of 235 U fission chambers. Here we employed graphite and beryllium with a ratio of Be/C=0.25 as moderator, which materials are stable in ITER relevant temperature in a horizontal port. Based on the neutronics calculations, a fission chamber with 200 mg of 235 U is adopted for the neutron flux monitor. Three detectors are mounted in a stainless steel housing with moderation material. Two fission chamber assemblies will be installed in a horizontal port; one is for D-D and calibration operation, and another is for D-T operation. The assembly for the D-D operation and the calibration are installed just outside the port plug in the horizontal port. The assembly for the D-T operation is installed just behind the additional shield in the port. Combining of those assemblies with both pulse counting mode and Campbelling mode in the electronics, a dynamic range of 10 7 can be obtained with 1 ms temporal resolution. Effects of gamma-rays and magnetic fields on the fission chamber are negligible in this arrangement. The neutron flux monitor can meet the required 10% accuracy for a fusion power monitor. (author)

  16. Design of a cryogenic deuterium gas target for neutron therapy

    International Nuclear Information System (INIS)

    Kuchnir, F.T.; Waterman, F.M.; Forsthoff, H.; Skaggs, L.S.; Vander Arend, P.C.; Stoy, S.

    1976-01-01

    A cryogenic deuterium gas target operating at 80 0 K and 10 atm pressure has been designed for use with a small cyclotron; the D(d,n) reaction is used to produce a neutron beam suitable for radiation therapy. The target is cooled by circulation of the gas in a closed loop between the target and an external heat exchanger immersed in liquid nitrogen

  17. Neutronics Study on LEU Nuclear Thermal Rocket Fuel Options

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Howe, Steven [CSNR, Idaho (United States)

    2014-10-15

    This has resulted in a non-trivial simplification of the tasks needed to develop such an engine and the quick initial development of the concept. There are, however, a series of key core-design choices that are currently under scrutiny in the field that have to be resolved in order for the LEU-NTR to be fully developed. The most important of these is the choice of fuel: carbide composite or tungsten cermet. This study presents a first comparison of the two fuel types specifically in the neutronic application to the LEU-NTR, keeping in mind the unique neutronic environment and the system requirements of the system. The scope of the study itself is limited to a neutronics study of the two fuels and only a cursory overview of the material properties of the fuels themselves... The results of this study have led to two major conclusions. First of all is that the carbide composite fuel is, from a neutronics standpoint, a much better fuel. It has a low absorption cross-section, is inherently a strong moderator, is able to achieve a higher reactivity using smaller amounts of fissile material, and can potentially enable a smaller reactor. Second is that despite its neutronic difficulties (high absorption, inferior moderating abilities, and lower k-infinity values) the tungsten cermet fuel is still able to perform satisfactorily in an LEU-NTR, largely due to its ability to have an extremely high fuel loading.

  18. Modernization design of neutron radiography of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, B.; Bilge, A.N.

    1988-01-01

    ITU TRIGA MARK-II Research and Training Reactor has a power of 250 KW and has three beam tubes. One of them is tangential beam tube used for neutron radiography. In this study, the neutron radiography set in the tangential beam tube is described with its problems for ITU TRIGA Reactor. After that modernization of the system is designed and the applicability of the direct and indirect methods is evaluated. Improving the ratio of length to diameter for the beam tube, elimination the fogging on the film and constructive design for practice and secure application of the technique is developed. (author)

  19. The Spallation Neutron Source (SNS) conceptual design shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Odano, N.; Lillie, R.A.

    1998-03-01

    The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

  20. A proposed experiment for studying the direct neutron-neutron interaction

    International Nuclear Information System (INIS)

    Hassan Fikry, A.R.; Maayouf, R.M.A.

    1979-01-01

    An experiment for studying the direct neutron-neutron interaction is suggested. The experiment is based on the combined use of an accelerator, e.g., an electron linear accelerator, together with a mobile pulsed reactor; or using a pulsed beam reactor together with a mobile neutron generator

  1. 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

  2. Neutron detector array at IUAC: Design features and instrumentation ...

    Indian Academy of Sciences (India)

    2014-11-02

    Nov 2, 2014 ... multi-wire proportional counters mounted inside a 1 m diameter SS target chamber. ... a powerful tool to study the dynamics of the fissioning system and a ... neutron multiplicity for three isotopes across a major closed shell, .... voltage by connecting to the control pin of the chip via a 0–5V programmable volt-.

  3. Neutron scattering studies of Mn12-acetate

    International Nuclear Information System (INIS)

    Robinson, R.A.

    2000-01-01

    Full text: The S=10 magnetic molecule Mn 12 -acetate, which crystallises into a tetragonal crystal structure, has attracted substantial recent attention by virtue of its low temperature bulk magnetic properties, which give evidence for resonant quantum tunnelling of the magnetisation. We report a full neutron crystal structure including positions of all protons/deuterons, including the solvated water and acetic acid, a polarised-neutron study of the real space magnetisation, which confirms a simple magnetic-structure model for the molecule, albeit with reduced Mn moments, and inelastic neutron scattering data containing both the excitations within the 21-fold degenerate S=10 manifold, and those from S=10 to the S=9 manifolds. Both manifolds are split by uniaxial magnetic anisotropy, and we report coefficients for 2nd and 4th-order terms in the magnetic Hamiltonian

  4. Fusion materials high energy-neutron studies. A status report

    International Nuclear Information System (INIS)

    Doran, D.G.; Guinan, M.W.

    1980-01-01

    The objectives of this paper are (1) to provide background information on the US Magnetic Fusion Reactor Materials Program, (2) to provide a framework for evaluating nuclear data needs associated with high energy neutron irradiations, and (3) to show the current status of relevant high energy neutron studies. Since the last symposium, the greatest strides in cross section development have been taken in those areas providing FMIT design data, e.g., source description, shielding, and activation. In addition, many dosimetry cross sections have been tentatively extrapolated to 40 MeV and integral testing begun. Extensive total helium measurements have been made in a variety of neutron spectra. Additional calculations are needed to assist in determining energy dependent cross sections

  5. Spallation neutron source target design for radioactive waste transmutation

    International Nuclear Information System (INIS)

    Beard, C.A.

    1992-01-01

    The disposal of high-level radioactive waste has long been one of the most serious problems facing the nuclear industry. Transmutation of this waste through particle bombardment has been suggested numerous times as a possible method of enhancing the waste management process. Due to advances in accelerator technology, the feasibility of an accelerator based transmutation system has increased enough to allow serious investigation of this process. Therefore, in pursuit of this goal, an accelerator target was designed for use in an accelerator based transmutation system. The target design consists of an array of tantalum rods, cooled by liquid sodium, which are arranged in a cylindrical configuration 40 cm in diameter and 125 cm in height. Tantalum was chosen as the target material over tungsten, lead, bismuth, and a lead-bismuth alloy (55 w/o bismuth) due to a large neutron yield, low activation, low chemical toxicity, and the fact that it does not produce significant amounts of long-lived isotopes through spallation or activation. The target yields a neutron source of 29.7 neutrons/proton when exposed to a 1600 MeV proton beam, and is suitable for use with both thermal or fast spectrum transmutation systems

  6. Neutronic design of a traveling wave reactor core

    International Nuclear Information System (INIS)

    Lopez S, R. C.; Francois L, J. L.

    2010-10-01

    The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

  7. Neutron diffraction technique as a method for material studies

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2008-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  9. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2009-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  10. Neutron-scattering studies of magnetic superconductors

    International Nuclear Information System (INIS)

    Sinha, S.K.; Crabtree, G.W.; Hinks, D.G.; Mook, H.A.; Pringle, O.A.

    1982-01-01

    Results obtained in the last few years obtained by neutron diffraction on the nature of the magnetic ordering in magnetic superconductors are reviewed. Emphasis is given to studies of the complex intermediate phase in ferromagnetic superconductors where both superconductivity and ferromagnetism appear to coexist

  11. NEUTRON-SCATTERING STUDY OF DCN

    DEFF Research Database (Denmark)

    Mackenzie, Gordon A.; Pawley, G. S.

    1979-01-01

    Phonons in deuterium cyanide have been measured by neutron coherent inelastic scattering. The main subject of study was the transverse acoustic mode in the (110) direction polarised along (110) which is associated with the first-order structural phase transition at 160K. Measurements have shown...

  12. A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Nina Fauziah

    2015-03-01

    Full Text Available Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1. Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria   Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT di Reaktor Riset Kartini dengan menggunakan program Monte

  13. Design, construction and characterization of a new neutron beam for neutron radiography at the Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choopan Dastjerdi, M.H., E-mail: mdastjerdi@aeoi.org.ir [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Department of Energy Engineering and Physics, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of); Khalafi, H.; Kasesaz, Y.; Mirvakili, S.M.; Emami, J.; Ghods, H.; Ezzati, A. [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of)

    2016-05-11

    To obtain a thermal neutron beam for neutron radiography applications, a neutron collimator has been designed and implemented at the Tehran Research Reactor (TRR). TRR is a 5 MW open pool light water moderated reactor with seven beam tubes. The neutron collimator is implemented in the E beam tube of the TRR. The design of the neutron collimator was performed using MCNPX Monte Carlo code. In this work, polycrystalline bismuth and graphite have been used as a gamma filter and an illuminator, respectively. The L/D parameter of the facility was chosen in the range of 150–250. The thermal neutron flux at the image plane can be varied from 2.26×10{sup 6} to 6.5×10{sup 6} n cm{sup −2} s{sup −1}. Characterization of the beam was performed by ASTM standard IQI and foil activation technique to determine the quality of neutron beam. The results show that the obtained neutron beam has a good quality for neutron radiography applications.

  14. Preliminary design of GDT-based 14 MeV neutron source

    International Nuclear Information System (INIS)

    Du Hongfei; Chen Dehong; Wang Hui; Wang Fuqiong; Jiang Jieqiong; Wu Yican; Chen Yiping

    2012-01-01

    To meet the need of D-T fusion neutron source for fusion material testing, design goals were presented in this paper according to the international requirements of neutron source for fusion material testing. A preliminary design scheme of GDT-based 14 MeV neutron source was proposed, and a physics model of the neutron source was built based on progress of GDT experiments. Two preliminary design schemes (i. e. FDS-GDT1, FDS-GDT2) were designed; among which FDS-GDT2 can be used for fusion material testing with neutron first wall loading of 2 MW/m 2 . (authors)

  15. A study on the utilization of hyper-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1993-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwellian distribution of a higher temperature than the room temperature of 300 K, was studied in order to improve the thermal neutron flux distribution at the deeper part in a living body for neutron capture therapy. Simulation calculations were carried out using MCNP-V3 in order to confirm the characteristics of hyper-thermal neutrons, i.e., (1) depth dependence of neutron energy spectrum, and (2) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that the hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper and wider area in a living body compared with the thermal neutron irradiation. Practically, by the incidence of the hyper-thermal neutrons with a 3000 K Maxwellian distribution, the thermal neutron flux at 5 cm depth can be given about four times larger than by the incidence of the thermal neutrons of 300 K. (author)

  16. Virtual design of the neutron guide for the TOF spectrometer NEAT

    International Nuclear Information System (INIS)

    Izaola, Zunbeltz; Russina, Margarita

    2010-01-01

    We present the results of a virtual design study based on Monte-Carlo neutron ray tracing techniques for the neutron guide of the time of flight (TOF) spectrometer NEAT. We studied several configurations with linearly or elliptically tapered compressors with different degrees of focusing and different guide coatings. The calculations were performed and crosschecked using two software packages which produced similar results. The geometrical arrangement of selected guide components was optimised with the Particle Swarm Optimisation algorithm. The results of the Monte Carlo simulations confirm an expected intensity gain factor of approximately 5, that can be achieved by the optimal configuration.

  17. Virtual design of the neutron guide for the TOF spectrometer NEAT

    Science.gov (United States)

    Izaola, Zunbeltz; Russina, Margarita

    2010-11-01

    We present the results of a virtual design study based on Monte-Carlo neutron ray tracing techniques for the neutron guide of the time of flight (TOF) spectrometer NEAT. We studied several configurations with linearly or elliptically tapered compressors with different degrees of focusing and different guide coatings. The calculations were performed and crosschecked using two software packages which produced similar results. The geometrical arrangement of selected guide components was optimised with the Particle Swarm Optimisation algorithm. The results of the Monte Carlo simulations confirm an expected intensity gain factor of approximately 5, that can be achieved by the optimal configuration.

  18. Epithermal neutron beam design for neutron capture therapy at the Power Burst Facility and the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Wheeler, F.J.; Parsons, D.K.; Rushton, B.L.; Nigg, D.W.

    1990-01-01

    Nuclear design studies have been performed for two reactor-based epithermal neutron beams for cancer treatment by neutron capture therapy (NCT). An intermediate-intensity epithermal beam has been designed and implemented at the Brookhaven Medical Research Reactor (BMRR). Measurements show that the BMRR design predictions for the principal characteristics of this beam are accurate. A canine program for research into the biological effects of NCT is now under way at BMRR. The design for a high-intensity epithermal beam with minimal contamination from undesirable radiation components has been finalized for the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. This design will be implemented when it is determined that human NCT trials are advisable. The PBF beam will exhibit approximately an order of magnitude improvement in absolute epithermal flux intensity over that available in the BMRR, and its angular distribution and spectral characteristics will be more advantageous for NCT. The combined effects of beam intensity, angular distribution, spectrum, and contaminant level allow the desired tumor radiation dose to be delivered in much shorter times than are possible with the currently available BMRR beam, with a significant reduction (factor of 3 to 5) in collateral dose due to beam contaminants

  19. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6 Li in order to reduce parasite neutron captures in there. (orig./HP)

  20. Study of material science by neutron scattering

    International Nuclear Information System (INIS)

    Kim, H.J.; Yoon, B.K.; Cheon, B.C.; Lee, C.Y.; Kim, C.S.

    1980-01-01

    To develop accurate methods of texture measurement in metallic materials by neutron diffraction, (100),(200),(111) and (310) pole figures have been measured for the oriented silicon steel sheet, and currently study of correction methods for neutron absorption and extinction effects are in progress. For quantitative analysis of texture of polycrystalline material with a cubic structure, a software has been developed to calculate inverse pole figures for arbitrary direction specified in the speciman as well as pole figures for arbitrary chosen crystallographic planes from three experimental pole figures. This work is to be extended for the calculation of three dimensional orientation distribution function and for the evaluation of errors in the quantitative analysis of texture. Work is also for the study of N-H...O hydrogen bond in amino acid by observing molecular motions using neutron inelastic scattering. Measurement of neutron inelastic scattering spectrum of L-Serine is completed at 100 0 K and over the energy transfer range of 20-150 meV. (KAERI INIS Section)

  1. Kriging-based algorithm for nuclear reactor neutronic design optimization

    International Nuclear Information System (INIS)

    Kempf, Stephanie; Forget, Benoit; Hu, Lin-Wen

    2012-01-01

    Highlights: ► A Kriging-based algorithm was selected to guide research reactor optimization. ► We examined impacts of parameter values upon the algorithm. ► The best parameter values were incorporated into a set of best practices. ► Algorithm with best practices used to optimize thermal flux of concept. ► Final design produces thermal flux 30% higher than other 5 MW reactors. - Abstract: Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.

  2. Studies and modeling of cold neutron sources; Etude et modelisation des sources froides de neutron

    Energy Technology Data Exchange (ETDEWEB)

    Campioni, G

    2004-11-15

    With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources.

  3. A study of television imaging system for fast neutron radiography

    International Nuclear Information System (INIS)

    Yoshii, Koji

    1992-01-01

    The neutron radiography with fast neutron beam is a very useful imaging technique for thicker objects, especially those composed of hydrogen-rich materials which are sometimes difficult to image by thermal neutron radiography. The fast neutron radiography has not been studied so much as the thermal neutron radiography. The fast neutron radiography has been studied at the fast neutron source reactor 'Yayoi' of the University of Tokyo built in Tokai-mura. The average neutron energy of the Yayoi is about 1 MeV, and the peak neutron flux at the core center is 0.8 x 10 12 at the maximum operating power of 2 kW. In the experiment on fast neutron radiography, a CR39 nuclear track detector has been used successfully. But in the Yayoi radiography procedure, about 24 hours were required for obtaining an imaging result. To get a prompt imaging result and a real-time imaging result, it is necessary to develop a fast neutron television system, and in this paper, a new fast neutron TV system is proposed. The main difference is the converter material sensitive to fast neutrons. The study on the fast neutron TV system was carried out by using the Baby Cyclotron of Japan Steel Works, and the good images were realized. (K.I.)

  4. The advanced neutron source facility: Safety philosophy and studies

    International Nuclear Information System (INIS)

    Greene, S.R.; Harrington, R.M.

    1988-01-01

    The Advanced Neutron Source (ANS) is currently the only new civilian nuclear reactor facility proposed for construction in the United States. Even though the thermal power of this research-oriented reactor is a relatively low 300 MW, the design will undoubtedly receive intense scrutiny before construction is allowed to proceed. Safety studies are already under way to ensure that the maximum degree of safety in incorporated into the design and that the design is acceptable to the Department of Energy (DOE) and can meet the Nuclear Regulatory Commission regulations. This document discusses these safety studies

  5. A preliminary neutron crystallographic study of thaumatin

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Susana C. M. [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); Blakeley, Matthew P. [Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); Leal, Ricardo M. F. [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); ESRF, 6 Rue Jules Horowitz, BP-220, 38043 Grenoble (France); Mitchell, Edward P. [EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); ESRF, 6 Rue Jules Horowitz, BP-220, 38043 Grenoble (France); Forsyth, V. Trevor, E-mail: tforsyth@ill.fr [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom)

    2008-05-01

    Preliminary neutron crystallographic data from the sweet protein thaumatin have been recorded using the LADI-III diffractometer at the Institut Laue Langevin (ILL). The results illustrate the feasibility of a full neutron structural analysis aimed at further understanding the molecular basis of the perception of sweet taste. Such an analysis will exploit the use of perdeuterated thaumatin. A preliminary neutron crystallographic study of the sweet protein thaumatin is presented. Large hydrogenated crystals were prepared in deuterated crystallization buffer using the gel-acupuncture method. Data were collected to a resolution of 2 Å on the LADI-III diffractometer at the Institut Laue Langevin (ILL). The results demonstrate the feasibility of a full neutron crystallographic analysis of this structure aimed at providing relevant information on the location of H atoms, the distribution of charge on the protein surface and localized water in the structure. This information will be of interest for understanding the specificity of thaumatin–receptor interactions and will contribute to further understanding of the molecular mechanisms underlying the perception of taste.

  6. Current-mode detector for neutron time-of-flight studies

    International Nuclear Information System (INIS)

    Bowman, J.D.; Szymanski, J.J.; Yuan, V.W.; Bowman, C.D.; Silverman, A.; Zhu, X.

    1990-01-01

    A system for the detection of high-intensity neutron bursts with instantaneous rates as high as 10 11 Hz is presented. This system uses a transient digitizer to sample the output voltage of a high-current photomultiplier tube as a function of time. The coutput voltage is proportional to the number of neutrons striking the detector. This detector is used at the Los Alamos Neutron Scattering Center to study fundamental symmetries. Design considerations, construction details and performance benchmarks are presented. (orig.)

  7. Neutron scattering studies of modulated magnetic structures

    Energy Technology Data Exchange (ETDEWEB)

    Aagaard Soerensen, Steen

    1999-08-01

    This report describes investigations of the magnetic systems DyFe{sub 4}Al{sub 8} and MnSi by neutron scattering and in the former case also by X-ray magnetic resonant scattering. The report is divided into three parts: An introduction to the technique of neutron scattering with special emphasis on the relation between the scattering cross section and the correlations between the scattering entities of the sample. The theoretical framework of neutron scattering experiments using polarized beam technique is outlined. The second part describes neutron and X-ray scattering investigation of the magnetic structures of DyFe{sub 4}Al{sub 8}. The Fe sublattice of the compound order at 180 K in a cycloidal structure in the basal plane of the bct crystal structure. At 25 K the ordering of the Dy sublattice shows up. By the element specific technique of X-ray resonant magnetic scattering, the basal plane cycloidal structure was also found for the Dy sublattice. The work also includes neutron scattering studies of DyFe{sub 4}Al{sub 8} in magnetic fields up to 5 T applied along a <110> direction. The modulated structure at the Dy sublattice is quenched by a field lower than 1 T, whereas modulation is present at the Fe sublattice even when the 5 T field is applied. In the third part of the report, results from three small angle neutron experiments on MnSi are presented. At ambient pressure, a MnSi is known to form a helical spin density wave at temperature below 29 K. The application of 4.5 kbar pressure intended as hydrostatic decreased the Neel temperature to 25 K and changed the orientation of the modulation vector. To understand this reorientation within the current theoretical framework, anisotropic deformation of the sample crystal must be present. The development of magnetic critical scattering with an isotropic distribution of intensity has been studied at a level of detail higher than that of work found in the literature. Finally the potential of a novel polarization

  8. Design of the Mechanical Parts for the Neutron Guide System at HANARO

    International Nuclear Information System (INIS)

    Shin, J. W.; Cho, Y. G.; Cho, S. J.; Ryu, J. S.

    2008-01-01

    The research reactor HANARO (High-flux Advanced Neutron Application ReactOr) in Korea will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. Functions of the in-pile plug assembly are to shield the reactor environment from nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical structure to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This paper describes the design of the in-pile assembly and the primary shutter for the neutron guide system at HANARO. The design of the guide shielding assembly for the primary shutter and the neutron guides is also presented

  9. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    Cantargi, F; Sbaffoni, M; Granada, R

    2004-01-01

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed [es

  10. Utilization of low voltage D-T neutron generators in neutron physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Singkarat, S.

    1995-08-01

    In a small nuclear laboratory of a developing country a low voltage D-T neutron generator can be a very useful scientific apparatus. Such machines have been used successfully for more than 40 years in teaching and scientific research. The original continuous mode 150-kV D-T neutron generator has been modified to have also a capability of producing 2-ns pulsed neutrons. Together with a carefully designed 10 m long flight path collimator and shielding of a 25 cm diameter {center_dot} 10 cm thick BC-501 neutron detector, the pulsing system was successfully used for measuring the double differential cross-section (DDX) of natural iron for 14.1-MeV neutron from the angle of 30 deg to 150 deg in 10 deg steps. In order to extend the utility of the generator, two methods for converting the almost monoenergetic 14-MeV neutrons to monoenergetic neutrons of lower energy were proposed and tested. The first method uses a pulsed neutron generator and the second method uses an ordinary continuous mode generator. The latter method was successfully used to measure the scintillation light output of a 1.4 cm diameter spherical NE-213 scintillation detector. The neutron generator has also been used in the continuous search for improved neutron detection techniques. There is a proposal, based on Monte Carlo calculations, of using a scintillation fiber for a fast neutron spectrometer. Due to the slender shape of the fiber, the pattern of produced light gives a peak in the pulse height spectrum instead of the well-known rectangular-like distribution, when the fiber is bombarded end-on by a beam of 14-MeV neutrons. Experimental investigations were undertaken. Detailed investigations on the light transportation property of a short fiber were performed. The predicted peak has not yet been found but the fiber detector may be developed as a directional discrimination fast neutron detector. 18 refs.

  11. Design and implementation of the control system for neutron reflectometer

    International Nuclear Information System (INIS)

    Wu Xuehui; Fu Yongli; Zhou Aiyu; Zhu Kejun; Yuan Guangcui

    2011-01-01

    The neutron reflectometry is an important technique that has widespread applications as a powerful analytical tool to analyze the surface and interfacial structure and composition of many materials. An efficient and accurate instrument control system is a key component of the system, with software based on LabVIEW and hardware based on PCI-1240 motor control card, TRUMP-PCI-2K multichannel buffer card and 974 counter/timer. It gives an overview of the design and implementation of this control system. The results prove that this system fulfills the needs well with high stability and operability. (authors)

  12. THERMAL: A routine designed to calculate neutron thermal scattering

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1995-01-01

    THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy

  13. Target designs for the Brookhaven National Laboratory 5-MW pulsed spallation neutron source

    International Nuclear Information System (INIS)

    Ludewig, H.; Todosow, M.; Powell, J.R.

    1996-01-01

    A feasibility study of a compact high power density target for a spallation neutron source was under-taken. The target arrangement consists primarily of heavy metal, with appropriate cooling passages. A high intensity proton beam of intermediate energy is directed at the target, where it interacts with the heavy metal nuclei. The subsequent spallation reactions produce several neutrons per proton resulting in an intense neutron source. The proton beam is assumed to havean energy of 5 MW, and to be cyclic with a repetition rate of 10Hz and 50Hz. The study was divided into two broad sections. First, an analysis of preliminary target designs was undertaken to ensure the overall feasibility of the concepts involved in the design and eventual construction of such a high power density target. Second, two proposed target designs, based on the first set of analyses, are investigated in more detail. Special care is taken to ensure that the neutron fluxes in the moderator are at the desired level no material compatibility problems exist,and the target is able to operate in a reliable and safe manner. Several target materials, coolant types, and target arrangements are investigated in the first section. The second section concentrates on a single target material and geometric arrangement. However, several structural material choices continue to be investigated with the aim of minimizing the effects of structural heating, and associated thermally induced stresses. In the final section the conclusions of this preliminary study are summarized

  14. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  15. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  16. Neutron scattering studies of the actinides

    International Nuclear Information System (INIS)

    Lander, G.H.

    1979-01-01

    The electronic structure of actinide materials presents a unique example of the interplay between localized and band electrons. Together with a variety of other techniques, especially magnetization and the Mossbauer effect, neutron studies have helped us to understand the systematics of many actinide compounds that order magnetically. A direct consequence of the localization of 5f electrons is the spin-orbit coupling and subsequent spin-lattice interaction that often leads to strongly anisotropic behavior. The unusual phase transition in UO 2 , for example, arises from interactions between quadrupole moments. On the other hand, in the monopnictides and monochalcogenides, the anisotropy is more difficult to understand, but probably involves an interaction between actinide and anion wave functions. A variety of neutron experiments, including form-factor studies, critical scattering and measurements of the elementary excitations have now been performed, and the conceptual picture emerging from these studies will be discussed

  17. Annual report on neutron scattering studies in JAERI

    International Nuclear Information System (INIS)

    Sato, Masatoshi; Nishi, Masakazu; Fujishita, Hideshi; Iizumi, Masashi

    1982-07-01

    Neutron scattering studies carried out from September 1979 to August 1981 by Division of Physics, JAERI, and universities with JRR-2 and -3 neutron beam facilities are described: 61 summary reports, and a list of publications. (author)

  18. Conceptual design of 30 MeV magnet system used for BNCT epithermal neutron source

    International Nuclear Information System (INIS)

    Slamet Santosa; Taufik

    2015-01-01

    Conceptual design of 30 MeV Magnet System Used for BNCT Epithermal Neutron Source has been done based on methods of empirical model of basic equation, experiences of 13 MeV cyclotron magnet design and personal communications. In the field of health, cyclotron can be used as an epithermal neutron source for Boron Neutron Capture Therapy (BNCT). The development of cyclotron producing epithermal neutrons for BNCT has been performed at Kyoto University, of which it produces a proton beam current of 1.1 mA with energy of 30 MeV. With some experiences on 13 MeV cyclotron magnet design, to support BNCT research and development we performed the design studies of 30 MeV cyclotron magnet system, which is one of the main components of the cyclotron for deflecting proton beam into circular trajectory and serves as beam focusing. Results of this study are expected to define the parameters of particular cyclotron magnet. The scope of this study includes the study of the parameters component of the 30 MeV cyclotron and magnet initial parameters. The empirical method of basic equation model is then corroborated by a simulation using Superfish software. Based on the results, a 30 MeV cyclotron magnet for BNCT neutron source enables to be realized with the parameters of B 0 = 1.06 T, frequency RF = 64.733938 ≈ 65 MHz, the external radius of 0.73 m, the radius of the polar = 0.85 m, BH = 1.95 T and a gap hill of 4 cm. Because proton beam current that be needed for BNCT application is very large, then in the calculation it is chosen a great focusing axial νz = 0.630361 which can generate B V = 0.44 T. (author)

  19. The design of systems for the determination of plutonium by passive neutron counting

    International Nuclear Information System (INIS)

    Hooton, B.W.

    1978-10-01

    The properties of moderators and other materials commonly used in systems for determination of plutonium by passive neutron counting have been investigated. The neutron flux from spontaneous fission and (α,n) reactions has been evaluated and the design characteristics of a number of systems have been determined by Monte Carlo tracking of neutrons. (author)

  20. DNA hydration studied by neutron fiber diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, W.; Forsyth, V.T.; Mahendrasingam, A.; Langan, P.; Pigram, W.J. [Keele Univ. (United Kingdom)] [and others

    1994-12-31

    The development of neutron high angle fiber diffraction to investigate the location of water around the deoxyribonucleic acid (DNA) double-helix is described. The power of the technique is illustrated by its application to the D and A conformations of DNA using the single crystal diffractometer, D19, at the Institute Laue-Langevin, Grenoble and the time of flight diffractometer, SXD, at the Rutherford Appleton ISIS Spallation Neutron Source. These studies show the existence of bound water closely associated with the DNA. The patterns of hydration in these two DNA conformations are quite distinct and are compared to those observed in X-ray single crystal studies of two-stranded oligodeoxynucleotides. Information on the location of water around the DNA double-helix from the neutron fiber diffraction studies is combined with that on the location of alkali metal cations from complementary X-ray high angle fiber diffraction studies at the Daresbury Laboratory SRS using synchrotron radiation. These analyses emphasize the importance of viewing DNA, water and ions as a single system with specific interactions between the three components and provide a basis for understanding the effect of changes in the concentration of water and ions in inducing conformations] transitions in the DNA double-helix.

  1. DNA hydration studied by neutron fiber diffraction

    International Nuclear Information System (INIS)

    Fuller, W.; Forsyth, V.T.; Mahendrasingam, A.; Langan, P.; Pigram, W.J.

    1994-01-01

    The development of neutron high angle fiber diffraction to investigate the location of water around the deoxyribonucleic acid (DNA) double-helix is described. The power of the technique is illustrated by its application to the D and A conformations of DNA using the single crystal diffractometer, D19, at the Institute Laue-Langevin, Grenoble and the time of flight diffractometer, SXD, at the Rutherford Appleton ISIS Spallation Neutron Source. These studies show the existence of bound water closely associated with the DNA. The patterns of hydration in these two DNA conformations are quite distinct and are compared to those observed in X-ray single crystal studies of two-stranded oligodeoxynucleotides. Information on the location of water around the DNA double-helix from the neutron fiber diffraction studies is combined with that on the location of alkali metal cations from complementary X-ray high angle fiber diffraction studies at the Daresbury Laboratory SRS using synchrotron radiation. These analyses emphasize the importance of viewing DNA, water and ions as a single system with specific interactions between the three components and provide a basis for understanding the effect of changes in the concentration of water and ions in inducing conformations] transitions in the DNA double-helix

  2. Wines: water inelastic neutron scattering experimental study

    International Nuclear Information System (INIS)

    Risch, P.; Ait Abderrahim, H.; D'hondt, P.; Malabu, E.

    1997-01-01

    An intercomparison of calculated fast neutron flux (E > 1 MeV) traverse through a very thick water zone obtained using both S N , (DORT) and Monte-Carlo (TRIPOLI and MCBEND) codes in combination with different cross-sections libraries (based on ENDF/B-III, IV, V and VI), showed small discrepancies either between S N , and Monte-Carlo results or even between S N , or Monte-Carlo results when we consider different cross-sections libraries except for S N , calculation when using P 0 , cross-sections. In order to validate our calculations we looked for experimental data. Unfortunately no experiment, dedicated for the fast neutron transport in large thickness of water, was found in the literature. Therefore SCK-CEN and EDF decided to launch the WINES experiment which is dedicated to study this phenomenon. WINES sands for Water Inelastic Neutron scattering Experimental Study. The aim of this experiment is to provide-experimental data for validation of neutron transport codes and nuclear cross-sections libraries used for LWR surveillance dosimetry analysis. The experimental device is made of 1 m 3 cubic plexiglass container filled with demineralized water. At one face of this cube, a 235 U neutron fission source system is screwed. The source device is made of a 235 U (93 % weight enriched) 18.55 x 16 cm 2 plate cladded with aluminium which is inserted in neutron beam emerging from the graphite gas-cooled BR1 reactor. Fission chambers ( 238 U(n,f), 232 Th(n,f), 237 Np(n,f) and 235 U(n,f)) are used to measure the flux traverses on the central axis of the water cube perpendicular to the fission sources. In this paper we will compare the experimental data to the calculated results using the S N , transport code DORT with the P 3 , ELXSIR library, based on ENDF/B-V, and the P 7 -BUGLE-93 library, based on ENDF/B-VI as well as the Monte-Carlo transport code TRIPOLI with a cross-section library based on ENDF/B IV and ENDF/B-VI. (authors)

  3. Neutron reflectivity studies of ionomer blends

    CERN Document Server

    Gabrys, B J; Bucknall, D G; Vesely, D; Braiewa, R; Weiss, R A

    2002-01-01

    Preliminary results are presented of a neutron reflectivity study of the interfacial width between lithium- and zinc-sulphonated deuterated polystyrene with polycarbonate (PC). Both systems are partially miscible and exhibit an upper critical solution temperature behaviour. The interdiffusion in these systems was measured by annealing at a temperature above the glass-transition temperature of both polymers. The interfacial profiles obtained for these systems were described by symmetric Gaussian interfaces. No significant diffusion was observed. (orig.)

  4. Design of neutron beams at the Argonne Continuous Wave Linac (ACWL) for boron neutron capture therapy and neutron radiography

    International Nuclear Information System (INIS)

    Zhou, X.L.; McMichael, G.E.

    1994-01-01

    Neutron beams are designed for capture therapy based on p-Li and p-Sc reactions using the Argonne Continuous Wave Linac (ACWL). The p-Li beam will provide a 2.5 x 10 9 n/cm 2 s epithermal flux with 7 x 10 5 γ/cm 2 s contamination. On a human brain phantom, this beam allows an advantage depth (AD) of 10 cm, an advantage depth dose rate (ADDR) of 78 cGy/min and an advantage ratio (AR) of 3.2. The p-Sc beam offers 5.9 x 10 7 n/cm 2 s and a dose performance of AD = 8 cm and AR = 3.5, suggesting the potential of near-threshold (p,n) reactions such as the p-Li reaction at E p = 1.92 MeV. A thermal radiography beam could also be obtained from ACWL

  5. Feasibility study of the university of Utah TRIGA reactor power upgrade - Part I: Neutronics-based study in respect to control rod system requirements and design

    Directory of Open Access Journals (Sweden)

    Ćutić Avdo

    2013-01-01

    Full Text Available We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination with experimental procedures so as to assess the potential maximum University of Utah TRIGA power with the currently available control rod system and have created feasibility studies for assessing new core configurations that could provide higher core power levels. For the maximum determined power of a new University of Utah TRIGA core arrangement, a new control rod system was proposed.

  6. Neutron scattering study of dilute supercritical solutions

    International Nuclear Information System (INIS)

    Cochran, H.D.; Wignall, G.D.; Shah, V.M.; Londono, J.D.; Bienkowski, P.R.

    1994-01-01

    Dilute solutions in supercritical solvents exhibit interesting microstructures that are related to their dramatic macroscopic behavior. In typical attractive solutions, solutes are believed to be surrounded by clusters of solvent molecules, and solute molecules are believed to congregate in the vicinity of one another. Repulsive solutions, on the other hand, exhibit a local region of reduced solvent density around the solute with solute-solute congregation. Such microstructures influence solubility, partial molar volume, reaction kinetics, and many other properties. We have undertaken to observe these interesting microstructures directly by neutron scattering experiments on dilute noble gas systems including Ar. The three partial structure factors for such systems and the corresponding pair correlation functions can be determined by using the isotope substitution technique. The systems studied are uniquely suited for our objectives because of the large coherent neutron scattering length of the isotope 36 Ar and because of the accurate potential energy functions that are available for use in molecular simulations and theoretical calculations to be compared with the scattering results. We will describe our experiment, the unique apparatus we have built for it, and the neutron scattering results from our initial allocations of beam time. We will also describe planned scattering experiments to follow those with noble gases, including study of long-chain molecules in supercritical solvents. Such studies will involve hydrocarbon mixtures with and without deuteration to provide contrast

  7. Utilization of low voltage D-T neutron generators in neutron physics studies

    International Nuclear Information System (INIS)

    Singkarat, S.

    1995-01-01

    In a small nuclear laboratory of a developing country a low voltage D-T neutron generator can be a very useful scientific apparatus. Such machines have been used successfully for more than 40 years in teaching and scientific research. The original continuous mode 150-kV D-T neutron generator has been modified to have also a capability of producing 2-ns pulsed neutrons. Together with a carefully designed 10 m long flight path collimator and shielding of a 25 cm diameter · 10 cm thick BC-501 neutron detector, the pulsing system was successfully used for measuring the double differential cross-section (DDX) of natural iron for 14.1-MeV neutron from the angle of 30 deg to 150 deg in 10 deg steps. In order to extend the utility of the generator, two methods for converting the almost monoenergetic 14-MeV neutrons to monoenergetic neutrons of lower energy were proposed and tested. Both designs used the neutron-proton interaction at a circular surface-of-revolution made of hydrocarbon materials. The first design is for a pulsed neutron generator and the second design is for an ordinary continuous mode generator. The latter method was successfully used to measure the scintillation light output of a 1.4 cm diameter spherical NE-213 scintillation detector. The neutron generator has also been used in the continuous search for improved neutron detection techniques. There is a proposal, based on Monte Carlo calculations, of using a scintillation fiber for a fast neutron spectrometer. Due to the slender shape of the fiber, the pattern of produced light gives a peak in the pulse height spectrum instead of the well-known rectangular-like distribution, when the fiber is bombarded end-on by a beam of 14-MeV neutrons. Experimental investigations were undertaken. Detailed investigations on the light transportation property of a short fiber were performed. The predicted peak has not yet been found but the fiber detector may be developed as a directional discrimination fast neutron

  8. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  9. Analytic scattering kernels for neutron thermalization studies

    International Nuclear Information System (INIS)

    Sears, V.F.

    1990-01-01

    Current plans call for the inclusion of a liquid hydrogen or deuterium cold source in the NRU replacement vessel. This report is part of an ongoing study of neutron thermalization in such a cold source. Here, we develop a simple analytical model for the scattering kernel of monatomic and diatomic liquids. We also present the results of extensive numerical calculations based on this model for liquid hydrogen, liquid deuterium, and mixtures of the two. These calculations demonstrate the dependence of the scattering kernel on the incident and scattered-neutron energies, the behavior near rotational thresholds, the dependence on the centre-of-mass pair correlations, the dependence on the ortho concentration, and the dependence on the deuterium concentration in H 2 /D 2 mixtures. The total scattering cross sections are also calculated and compared with available experimental results

  10. Exchange bias studied with polarized neutron reflectivity

    International Nuclear Information System (INIS)

    Velthuis, S. G. E. te

    2000-01-01

    The role of Polarized Neutron Reflectivity (PNR) for studying natural and synthetic exchange biased systems is illustrated. For a partially oxidized thin film of Co, cycling of the magnetic field causes a considerable reduction of the bias, which the onset of diffuse neutron scattering shows to be due to the loosening of the ferromagnetic domains. On the other hand, PNR measurements of a model exchange bias junction consisting of an n-layered Fe/Cr antiferromagnetic (AF) superlattice coupled with an m-layered Fe/Cr ferromagnetic (F) superlattice confirm the predicted collinear magnetization in the two superlattices. The two magnetized states of the F (along or opposite to the bias field) differ only in the relative orientation of the F and adjacent AF layer. The possibility of reading clearly the magnetic state at the interface pinpoints the commanding role that PNR is having in solving this intriguing problem

  11. Leggett's noncontextual model studied with neutrons

    International Nuclear Information System (INIS)

    Durstberger-Rennhofer, K.; Sponar, S.; Badurek, G.; Hasegawa, Y.; Schmitzer, C.; Bartosik, H.; Klepp, J.

    2011-01-01

    Full text: It is a long-lasting debate whether nature can be described by deterministic hidden variable theories (HVT) underlying quantum mechanics (QM). Bell inequalities for local HVT as well as the Kochen- Specker theorem for non-contextual models stress the conflict between these alternative theories and QM. Leggett showed that even nonlocal hidden variable models are incompatible with quantum predictions. Neutron interferometry and polarimetry are very proper tools to analyse the behaviour of single neutron systems, where entanglement is created between different degrees of freedom (e.g., spin/ path, spin/energy) and thus quantum contextuality can be studied. We report the first experimental test of a contextual model of quantum mechanics a la Leggett, which deals with definiteness of measurement results before the measurements. The results show a discrepancy between our model and quantum mechanics of more than 7 standard deviations and confirm quantum indefiniteness under the contextual condition. (author)

  12. Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

    Energy Technology Data Exchange (ETDEWEB)

    Didi, Abdessamad; Dadouch, Ahmed; Tajmouati, Jaouad; Bekkouri, Hassane [Advanced Technology and Integration System, Dept. of Physics, Faculty of Science Dhar Mehraz, University Sidi Mohamed Ben Abdellah, Fez (Morocco); Jai, Otman [Laboratory of Radiation and Nuclear Systems, Dept. of Physics, Faculty of Sciences, Tetouan (Morocco)

    2017-06-15

    Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

  13. Considerations for boron neutron capture therapy studies

    International Nuclear Information System (INIS)

    Faria Gaspar, P. de.

    1994-01-01

    Radiotherapy is indispensable as a mean to eradicate deeply or infiltrating tumor tissue that can not be removed surgically. Therefore, it is not selective and may also kill the surrounding health tissue. The principle of BNCT (Boron Neutron Capture Therapy) consist in targeting a tumor selectively with a boron-10 compound. This nuclide has a large capture cross section for thermal neutrons and the nuclear reaction and the delivered energy in locus will selective the tumor. Since its initial proposal in 1963 BNCT has made much progress, however it is not used in a routine treatment. In this work it was approached some complex procedures, as the obtention of selective boron compounds, the adequate set up of neutron beams, the biodistribution, the in vivo and in vitro studies, and also human patients treatments. This work provide fundamentals about BNCT to professional of different areas of knowledge since it comprises multidisciplinary study. It includes appendixes for the ones not related to the field for a better comprehension of the many aspects involved. It is also presented a glossary containing technical and basic aspects involved. It is also presented a glossary containing technical and basic terms referred in the work. (author). 174 refs, 1 fig, 12 apps

  14. Studies on neutron detection with solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Khouri, M.C.; Vilela, E.C.; Andrade, C. de.

    1993-03-01

    The detection of thermal and fast neutrons was studied. For thermal neutrons, alpha sensitive plastic was used in order to register the products of nuclear reactions taking place in boron and /or lithium converters. Fast neutrons produce recoil tracks within the detector. In the present case, CR-39 and Makrofol E were used. Chemical and electrochemical etching processes were used for thermal and fast neutron detectors, respectively. (F.E.). 6 refs, 4 figs, 6 tabs

  15. Neutron and proton transmutation-activation cross section libraries to 150 MeV for application in accelerator-driven systems and radioactive ion beam target-design studies

    International Nuclear Information System (INIS)

    Koning, A.J.; Chadwick, M.B.; MacFarlane, R.E.; Mashnik, S.; Wilson, W.B.

    1998-05-01

    New transmutation-activation nuclear data libraries for neutrons and protons up to 150 MeV have been created. These data are important for simulation calculations of radioactivity, and transmutation, in accelerator-driven systems such as the production of tritium (APT) and the transmutation of waste (ATW). They can also be used to obtain cross section predictions for the production of proton-rich isotopes in (p,xn) reactions, for radioactive ion beam (RIB) target-design studies. The nuclear data in these libraries stem from two sources: for neutrons below 20 MeV, we use data from the European activation and transmutation file, EAF97; For neutrons above 20 MeV and for protons at all energies we have isotope production cross sections with the nuclear model code HMS-ALICE. This code applies the Monte Carlo Hybrid Simulation theory, and the Weisskopf-Ewing theory, to calculate cross sections. In a few cases, the HMS-ALICE results were replaced by those calculated using the GNASH code for the Los Alamos LA150 transport library. The resulting two libraries, AF150.N and AF150.P, consist of 766 nuclides each and are represented in the ENDF6-format. An outline is given of the new representation of the data. The libraries have been checked with ENDF6 preprocessing tools and have been processed with NJOY into libraries for the Los Alamos transmutation/radioactivity code CINDER. Numerous benchmark figures are presented for proton-induced excitation functions of various isotopes compared with measurements. Such comparisons are useful for validation purposes, and for assessing the accuracy of the evaluated data. These evaluated libraries are available on the WWW at: http://t2.lanl.gov/. 21 refs

  16. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

    2004-01-01

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  17. Population-based metaheuristic optimization in neutron optics and shielding design

    Energy Technology Data Exchange (ETDEWEB)

    DiJulio, D.D., E-mail: Douglas.DiJulio@esss.se [European Spallation Source ERIC, P.O. Box 176, SE-221 00 Lund (Sweden); Division of Nuclear Physics, Lund University, SE-221 00 Lund (Sweden); Björgvinsdóttir, H. [European Spallation Source ERIC, P.O. Box 176, SE-221 00 Lund (Sweden); Department of Physics and Astronomy, Uppsala University, SE-751 20 Uppsala (Sweden); Zendler, C. [European Spallation Source ERIC, P.O. Box 176, SE-221 00 Lund (Sweden); Bentley, P.M. [European Spallation Source ERIC, P.O. Box 176, SE-221 00 Lund (Sweden); Department of Physics and Astronomy, Uppsala University, SE-751 20 Uppsala (Sweden)

    2016-11-01

    Population-based metaheuristic algorithms are powerful tools in the design of neutron scattering instruments and the use of these types of algorithms for this purpose is becoming more and more commonplace. Today there exists a wide range of algorithms to choose from when designing an instrument and it is not always initially clear which may provide the best performance. Furthermore, due to the nature of these types of algorithms, the final solution found for a specific design scenario cannot always be guaranteed to be the global optimum. Therefore, to explore the potential benefits and differences between the varieties of these algorithms available, when applied to such design scenarios, we have carried out a detailed study of some commonly used algorithms. For this purpose, we have developed a new general optimization software package which combines a number of common metaheuristic algorithms within a single user interface and is designed specifically with neutronic calculations in mind. The algorithms included in the software are implementations of Particle-Swarm Optimization (PSO), Differential Evolution (DE), Artificial Bee Colony (ABC), and a Genetic Algorithm (GA). The software has been used to optimize the design of several problems in neutron optics and shielding, coupled with Monte-Carlo simulations, in order to evaluate the performance of the various algorithms. Generally, the performance of the algorithms depended on the specific scenarios, however it was found that DE provided the best average solutions in all scenarios investigated in this work.

  18. A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; An, Shigehiro.

    1983-01-01

    Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

  19. Study of scattering in bi-dimensional neutron radiographic images

    International Nuclear Information System (INIS)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, F.C.

    2009-01-01

    The effect of neutron scattering frequently causes distortions in neutron radiographic images and, thus, reduces the quality. In this project, a type of filter, comprised of cadmium (a neutron absorber), was used in the form of a grid to correct this effect. This device generated image data in the discrete shadow bands of the absorber, components relative to neutron scattering on the test object and surroundings. Scattering image data processing, together with the original neutron radiographic image, resulted in a corrected image with improved edge delineation and, thus, greater definition in the neutron radiographic image of the test object. The objective of this study is to propose a theoretical/experimental methodology that is capable of eliminating the components relative to neutron scattering in neutron radiographic images, coming from the material that composes the test object and the materials that compose the surrounding area. (author)

  20. Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

    2010-01-01

    This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

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

    International Nuclear Information System (INIS)

    Rony Djokorayono; Agus Cahyono

    2014-01-01

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

  2. TOF neutron diffraction study of archaeological ceramics

    International Nuclear Information System (INIS)

    Kockelmann, W.; Kirfel, A.

    1999-01-01

    Complete text of publication follows. The time-of flight (TOF) neutron diffractometer ROTAX [1] at ISIS has been used for identification and quantitative phase analysis of archaeological pottery. Neutron diffraction yields mineral phase fractions which, in parallel with information obtained from other archaeometric examination techniques, can provide a fingerprint that can be used to identify provenance and reconstruct methods of manufacturing of an archaeological ceramic product. Phase fractions obtained from a 13th century Rhenish stoneware jar compare well with those obtained from a powder sample prepared from the same fragment. This indicates that reliable results can be obtained by illuminating a large piece or even an intact ceramic object making TOF neutron diffraction a truly non-destructive examination technique. In comparison to X-ray diffraction, information from the bulk sample rather than from surface regions is obtained. ROTAX allows for a simple experimental set-up, free of sample movements. Programmes of archaeological study on ROTAX involve Russian samples (Upper-Volga culture, 5000-2000 BC), Greek pottery, (Agora/Athens, 500-300 BC), and medieval German earthenware and stoneware ceramics (Siegburg waster heap, 13-15th century). (author)

  3. Study of computerized tomography using neutron beam

    International Nuclear Information System (INIS)

    Pereira, W.W.

    1991-05-01

    This paper aims to demonstrate the advantages, shortcomings and complementaries of a tomography development using neutrons over the one employing gamma rays in the context of their applications to non destructive essays. A simulated experimental study was performed in order to compare the two aforementioned tomographic procedures as applied to some materials. These materials were chosen for their clear advantages and complementaries as, for instance, aluminium, iron, plastic and aluminium hydroxide. In this work two tomographic systems, are employed both with parallel beams. The first with a gamma radiation source (Caesium-137), with an energy of 662 KeV and an activity of 3,9 x 10 9 Bq (100 mCi) and the second one employing a neutron source, the Argonaut Reactor of the Instituto de Engenharia Nuclear, IEN/CNEN, from where the thermal neutron beam of about 10 5 n/(cm.s) was obtained. It is possible to conclude from the simulated and experimental results, by means of image analysis and distortion measurements, that for a given material the adequate radiation and its energy may be chosen so as to better characterize it. (author)

  4. Dense granular Flows: a conceptual design of high-power neutron source

    Directory of Open Access Journals (Sweden)

    Yang Lei

    2017-01-01

    Full Text Available A high-power neutron source system is very useful for multifunctional applications, such as material facilities for advanced nuclear power, space radiation studies, radiography and tomography. Here the idea of inclined dense granular flow is utilized and developed in a new conceptual design of a compact high-power target to produce a high-energy and high-flux neutron irradiation (the flux is up to 1015 n/cm2/s or even 1016. Comparing to the traditional solid and liquid heavy metal targets, this design has advantages in material choice, fluid stability, heat removal, etc. In this paper the natures of the granular flows in an inclined chute are investigated and preliminary experimental and numerical results are reported. Then the feasibility of this design is discussed.

  5. The Neutron Star Interior Composition Explorer (NICER): Design and Development

    Science.gov (United States)

    Gendreau, Keith C.; Arzoumanian, Zaven; Adkins, Phillip W.; Albert, Cheryl L.; Anders, John F.; Aylward, Andrew T.; Baker, Charles L.; Balsamo, Erin R.; Bamford, William A.; Benegalrao, Suyog S.; hide

    2016-01-01

    During 2014 and 2015, NASA's Neutron star Interior Composition Explorer (NICER) mission proceeded successfully through Phase C, Design and Development. An X-ray (0.2{12 keV) astrophysics payload destined for the International Space Station, NICER is manifested for launch in early 2017 on the Commercial Resupply Services SpaceX-11 flight. Its scientific objectives are to investigate the internal structure, dynamics, and energetics of neutron stars, the densest objects in the universe. During Phase C, flight components including optics, detectors, the optical bench, pointing actuators, electronics, and others were subjected to environmental testing and integrated to form the flight payload. A custom-built facility was used to co-align and integrate the X-ray \\concentrator" optics and silicon-drift detectors. Ground calibration provided robust performance measures of the optical (at NASA's Goddard Space Flight Center) and detector (at the Massachusetts Institute of Technology) subsystems, while comprehensive functional tests prior to payload-level environmental testing met all instrument performance requirements. We describe here the implementation of NICER's major subsystems, summarize their performance and calibration, and outline the component-level testing that was successfully applied.

  6. Design of ITER neutron monitor using micro fission chambers

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

    1998-08-01

    We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

  7. A neutron spectrometer for studying giant resonances with (p,n) reactions in inverse kinematics

    International Nuclear Information System (INIS)

    Stuhl, L.; Krasznahorkay, A.; Csatlós, M.; Algora, A.; Gulyás, J.; Kalinka, G.; Timár, J.; Kalantar-Nayestanaki, N.; Rigollet, C.; Bagchi, S.; Najafi, M.A.

    2014-01-01

    A neutron spectrometer, the European Low-Energy Neutron Spectrometer (ELENS), has been constructed to study exotic nuclei in inverse-kinematics experiments. The spectrometer, which consists of plastic scintillator bars, can be operated in the neutron energy range of 100 keV–10 MeV. The neutron energy is determined using the time-of-flight technique, while the position of the neutron detection is deduced from the time-difference information from photomultipliers attached to both ends of each bar. A novel wrapping method has been developed for the plastic scintillators. The array has a larger than 25% detection efficiency for neutrons of approximately 500 keV in kinetic energy and an angular resolution of less than 1°. Details of the design, construction and experimental tests of the spectrometer will be presented

  8. SU-F-T-183: Design of a Beam Shaping Assembly of a Compact DD-Based Boron Neutron Capture Therapy System

    Energy Technology Data Exchange (ETDEWEB)

    Hsieh, M; Liu, Y; Nie, L [Purdue University, West Lafayette, Indiana (United States)

    2016-06-15

    Purpose: To design a beam shaping assembly (BSA) to shape the 2.45-MeV neutrons produced by a deuterium-deuterium (DD) neutron generator and to optimize the beam output for boron neutron capture therapy of brain tumors Methods: MCNP is used for this simulation study. The simulation model consists of a neutron surface source that resembles an actual DD source and is surrounded by a BSA. The neutron source emits 2.45-MeV neutrons isotropically. The BSA is composed of a moderator, reflector, collimator and filter. Various types of materials and geometries are tested for each component to optimize the neutron output. Neutron characteristics are measured with an 2×2×2-cm{sup 3} air-equivalent cylinder at the beam exit. The ideal BSA is determined by evaluating the in-air parameters, which include epithermal neutron per source neutron, fast neutron dose per epithermal neutron, and photon dose per epithermal neutron. The parameter values are compared to those recommended by the IAEA. Results: The ideal materials for reflector and thermal neutron filter were lead and cadmium, respectively. The thickness for reflector was 43 cm and for filter was 0.5 mm. At present, the best-performing moderator has 25 cm of AlF{sub 3} and 5 cm of MgF{sub 2}. This layout creates a neutron spectrum that has a peak at approximately 10 keV and produces 1.35E-4 epithermal neutrons per source neutron per cm{sup 2}. Additional neutron characteristics, fast neutrons per epithermal neutron and photon per epithermal neutron, are still under investigation. Conclusion: Working is ongoing to optimize the final layout of the BSA. The neutron spectrum at the beam exit window of the final configuration will have the maximum number of epithermal neutrons and limited photon and fast neutron contaminations within the recommended values by IAEA. Future studies will also include phantom experiments to validate the simulation results.

  9. Design of a facility for the in situ measurement of catalytic reaction by neutron scattering spectroscopy

    Science.gov (United States)

    Tan, Shuai; Cheng, Yongqiang; Daemen, Luke L.; Lutterman, Daniel A.

    2018-01-01

    Catalysis is a critical enabling science for future energy needs. The next frontier of catalysis is to evolve from catalyst discovery to catalyst design, and for this next step to be realized, we must develop new techniques to better understand reaction mechanisms. To do this, we must connect catalytic reaction rates and selectivities to the kinetics, energetics, and dynamics of individual elementary steps and relate these to the structure and dynamics of the catalytic sites involved. Neutron scattering spectroscopies offer unique capabilities that are difficult or impossible to match by other techniques. The current study presents the development of a compact and portable instrumental design that enables the in situ investigation of catalytic samples by neutron scattering techniques. The developed apparatus was tested at the Spallation Neutron Source (SNS) in Oak Ridge National Laboratory and includes a gas handling panel that allows for computer hookups to control the panel externally and online measurement equipment such as coupled GC-FID/TCD (Gas Chromatography-Flame Ionization Detector/Thermal Conductivity Detector) and MS (Mass Spectrometry) to characterize offgassing while the sample is in the neutron scattering spectrometer. This system is flexible, modular, compact, and portable enabling its use for many types of gas-solid and liquid-solid reactions at the various beamlines housed at the SNS.

  10. Design of incoming neutron-beam for detecting oil dirt

    International Nuclear Information System (INIS)

    Zhao Jingwu; Chen Xiaocheng; Alimujiang Naimaiti; Aierken Abuliemu

    2012-01-01

    For the technique of neutron back-scattering, the neutron counts are non-linear and have a tendency toward saturation because of the neutron self-shielding. As a result, the measurement accuracy is reduced and the measurement range is limited. Using a simply model and comparing with experimental data, it is shown that, in the measurement of the thickness of oil dirt, by adjusting the ratio of thermal to epithermal neutrons, the neutron self: shielding is weakened. As a result, the non-linearity can be reduced and the measurement accuracy and range can be improved. (authors)

  11. Neutron beam-line shield design for the protein crystallography instrument at the Lujan Center

    International Nuclear Information System (INIS)

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

    2001-01-01

    We have developed a very useful methodology for calculating absolute total (neutron plus gamma-ray) dose equivalent rates for use in the design of neutron beam line shields at a spallation neutron source. We have applied this technique to the design of beam line shields for several new materials science instruments being built at the Manuel Lujan Jr. Neutron Scattering Center. These instruments have a variety of collimation systems and different beam line shielding issues. We show here some specific beam line shield designs for the Protein Crystallography Instrument. (author)

  12. Model-based design evaluation of a compact, high-efficiency neutron scatter camera

    Science.gov (United States)

    Weinfurther, Kyle; Mattingly, John; Brubaker, Erik; Steele, John

    2018-03-01

    This paper presents the model-based design and evaluation of an instrument that estimates incident neutron direction using the kinematics of neutron scattering by hydrogen-1 nuclei in an organic scintillator. The instrument design uses a single, nearly contiguous volume of organic scintillator that is internally subdivided only as necessary to create optically isolated pillars, i.e., long, narrow parallelepipeds of organic scintillator. Scintillation light emitted in a given pillar is confined to that pillar by a combination of total internal reflection and a specular reflector applied to the four sides of the pillar transverse to its long axis. The scintillation light is collected at each end of the pillar using a photodetector, e.g., a microchannel plate photomultiplier (MCP-PM) or a silicon photomultiplier (SiPM). In this optically segmented design, the (x , y) position of scintillation light emission (where the x and y coordinates are transverse to the long axis of the pillars) is estimated as the pillar's (x , y) position in the scintillator "block", and the z-position (the position along the pillar's long axis) is estimated from the amplitude and relative timing of the signals produced by the photodetectors at each end of the pillar. The neutron's incident direction and energy is estimated from the (x , y , z) -positions of two sequential neutron-proton scattering interactions in the scintillator block using elastic scatter kinematics. For proton recoils greater than 1 MeV, we show that the (x , y , z) -position of neutron-proton scattering can be estimated with < 1 cm root-mean-squared [RMS] error and the proton recoil energy can be estimated with < 50 keV RMS error by fitting the photodetectors' response time history to models of optical photon transport within the scintillator pillars. Finally, we evaluate several alternative designs of this proposed single-volume scatter camera made of pillars of plastic scintillator (SVSC-PiPS), studying the effect of

  13. 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.

  14. General features of the neutronics design code EQUICYCLE

    International Nuclear Information System (INIS)

    Jirlow, K.

    1978-10-01

    The neutronics code EQUICYCLE has been developed and improved over a long period of time. It is expecially adapted to survey type design calculations of large fast power reactors with particular emphasis on the nuclear parameters for a realistic equilibrium fuel cycle. Thus the code is used to evaluate the breeding performance, the power distributions and the uranium and plutonium mass balance for realistic refuelling schemes. In addition reactivity coefficients can be calculated and the influence of burnup could be assessed. The code is two-dimensional and treats the reactor core in R-Z geometry. The basic ideas of the calculating scheme are successive iterative improvement of cross-section sets and flux spectra and use of the mid-cycle flux for burning the fuel according to a specified refuelling scheme. Normally given peak burn-ups and maximum power densities are used as boundary conditions. The code is capable of handling the unconventional, so called heterogeneous cores. (author)

  15. Design, construction and description of a triple axis neutron crystal spectrometer

    International Nuclear Information System (INIS)

    Fulfaro, R.; Vinhas, L.A.; Fuhrmann, C.; Liguori Neto, R.; Parente, C.B.R.

    1977-01-01

    Integrating the neutron inelastic scattering plan in IEA (Sao Paulo, Brazil) it was completely designed and constructed there, a triple axis neutron spectrometer. The details about construction and design of the spectrometer are described. Basic principles about the experimental method which utilizes neutron coherent inelastic scattering in order to determine the relation dispersion between frequency and wave vector of the crystalline vibrations, are presented [pt

  16. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1975-01-01

    The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  17. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1974-01-01

    The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  18. Neutron emission study after muon capture by nuclei

    International Nuclear Information System (INIS)

    Bouyssy, Alain.

    1974-01-01

    Muon capture by nuclei, used in the beginning for checking the weak interaction, is now a method of investigation of nuclear structure. Study of spectrum, asymmetry and polarization of emitted neutrons after polarized muon capture has been done in three directions: weak coupling constants, final state interaction, nuclear wave functions. The neutron intensity and helicity are very dependent of the neutron - residual nucleus interaction, while the asymmetry is sensitive to the wave functions used for the proton. Moreover if the induced tensor coupling constant is different from zero the asymmetry is increased. Longitudinal polarization experiments, with those for neutron intensity, would be of great interest to give informations on neutron asymmetry [fr

  19. Zeolite function studied by neutron diffraction

    International Nuclear Information System (INIS)

    Newsam, J.M.

    1988-01-01

    Some recent figures relating to industrial uses of zeolites are summarized. Recent advances in the application of neutron diffraction to zeolite science are overviewed, with particular emphasis on powder diffraction (PND) results. Single crystal neutron diffraction studies of some 17 hydrated natural and synthetic zeolites have now appeared and they provide a consistent picture of zeolite-water interactions. Complete PND studies of hydrated synthetic ABW- and SOD-framework zeolites have also been reported. Other PND studies have explored the structural consequences of non-framework cation exchange, of framework modification by dealumination, and of framework cation substitution. Relatively simple zeolite-hydrocarbon sorbate complexes that have been studied include benzene in zeolite Y, and benzene and pyridine in zeolite L. Areas that are well poised for further development include further extensions to lower symmetry systems, the use of PND data for zeolite structure solution, studies at elevated temperatures and pressures, and further studies of zeolite sorbate complexes. (author) 68 refs., 7 figs

  20. 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)

  1. Study of an integrated electronic monitor for neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Barelaud, B.; Nexon-Mokhtari, F.; Barrau, C.; Decossas, J.L.; Vareille, J.C. [Limoges Univ., 87 (France); Sarrabayrouse, G.J. [CNRS, 31 - Toulouse (France)

    1995-12-31

    The majority of the individual neutron monitors measure incorrectly in a certain energy range (10 keV-500 keV). For electronic devices, this problem is due to their high {gamma} sensitivity. To solve this problem a microelectronic detector design for neutron spectrometry is now being studied at LEPOFI. It is based on the measurement of the total energy which is deposited by {sup 6}Li (n, {alpha}) {sup 3}H or {sup 10}B (n {alpha}) {sup 7}Li reactions in silicon detectors. A new electronic sensor - for example a boron or lithium sandwich device -has been developed in collaboration with LAAS. Specific techniques for silicon sensor coating with boron have been developed and are briefly presented. The response of the detector has been computed using a model and a code developed at LEPOFI. Several parameters have been taken into account in the calculations: the type of layer, its thickness, the characteristics of the detector. The results of these simulations are discussed and calculated pulse height distributions for various designs are presented. The present status of the investigation only concerns thermal neutrons. (author).

  2. Non-destructive studies of fuel pellets by neutron resonance absorption radiography and thermal neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California, Berkeley, CA 94720 (United States); Vogel, S.C.; Mocko, M.; Bourke, M.A.M.; Yuan, V.; Nelson, R.O.; Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

    2013-09-15

    Many isotopes in nuclear materials exhibit strong peaks in neutron absorption cross sections in the epithermal energy range (1–1000 eV). These peaks (often referred to as resonances) occur at energies specific to particular isotopes, providing a means of isotope identification and concentration measurements. The high penetration of epithermal neutrons through most materials is very useful for studies where samples consist of heavy-Z elements opaque to X-rays and sometimes to thermal neutrons as well. The characterization of nuclear fuel elements in their cladding can benefit from the development of high resolution neutron resonance absorption imaging (NRAI), enabled by recently developed spatially-resolved neutron time-of-flight detectors. In this technique the neutron transmission of the sample is measured as a function of spatial location and of neutron energy. In the region of the spectra that borders the resonance energy for a particular isotope, the reduction in transmission can be used to acquire an image revealing the 2-dimensional distribution of that isotope within the sample. Provided that the energy of each transmitted neutron is measured by the neutron detector used and the irradiated sample possesses neutron absorption resonances, then isotope-specific location maps can be acquired simultaneously for several isotopes. This can be done even in the case where samples are opaque or have very similar transmission for thermal neutrons and X-rays or where only low concentrations of particular isotopes are present (<0.1 atom% in some cases). Ultimately, such radiographs of isotope location can be utilized to measure isotope concentration, and can even be combined to produce three-dimensional distributions using tomographic methods. In this paper we present the proof-of-principle of NRAI and transmission Bragg edge imaging performed at Flight Path 5 (FP5) at the LANSCE pulsed, moderated neutron source of Los Alamos National Laboratory. A set of urania mockup

  3. CAREM 25: actual status of the core neutronic design. Calculation line

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author) [es

  4. Algorithm for the real-structure design of neutron supermirrors

    International Nuclear Information System (INIS)

    Pleshanov, N.K.

    2004-01-01

    The effect of structure imperfections of neutron supermirrors on their performance is well known. Nevertheless, supermirrors are designed with the algorithms based on the theories of reflection from perfect layered structures. In the present paper an approach is suggested, in which the design of a supermirror is made on the basis of its real-structure model (the RSD algorithm) with the use of exact numerical methods. It allows taking the growth laws and the reflectance of real structures into account. The new algorithm was compared with the Gukasov-Ruban-Bedrizova (GRB) algorithm and with the most frequently used algorithm of Hayter and Mook (HM). Calculations showed that, when the parameters of the algorithms are chosen so that the supermirrors designed for a given angular acceptance m have the same number of bilayers, (a) for perfect layers the GRB, HM and RSD algorithms generate sequences of practically the same reflectance; (b) for real structures with rough interfaces and interdiffusion the GRB and HM algorithms generate sequences with insufficient number of thinner layers and the RSD algorithm turns out to be more responsive and efficient. The efficiency of the RSD algorithm increases for larger m. In addition, calculations have been carried out to demonstrate the effect of fabrication errors and absorption on the reflectance of Ni/Ti supermirrors

  5. How the RRR neutronic characteristics impact on the mechanical design

    International Nuclear Information System (INIS)

    Villarino, E.; Coquibus, K.

    2005-01-01

    This paper describes how the neutronic characteristics of a very demanding research reactor facility impact on the mechanical design of the reactor core. The Replacement Research Reactor (RRR) for the Australian Nuclear Science and Technology Organization is described making emphasis in the mechanical solutions to improve the core performance. The compact core is located inside a chimney, surrounded by heavy water contained in the Reflector Vessel. The whole assembly is at the bottom of the Reactor Pool, which is full of de-mineralized light water acting as coolant and moderator and biological shielding. The core is an array of sixteen plate-type Fuel Assemblies (FAs) and five absorber plates, which are called Control Plates (CP). The coolant is light water, which flows upwards. The final design of the core layout and control guide boxes was adopted to minimize the flux and PPF perturbation during the normal operation. The four lateral control plates are used mainly to shutdown the reactor and to compensate large reactivity transients. The central and cruciform regulating plate is used to compensate the reactivity change during the cycle operation. The regulating plate does minimize perturbation on PPF and irradiation fluxes. The design of the reflector tank fulfills all the flux requirements for the irradiation facilities and also the flux perturbation between irradiation facilities. (authors)

  6. Advanced neutron source design: burnout heat flux correlation development

    International Nuclear Information System (INIS)

    Gambill, W.R.; Mochizuki, T.

    1988-01-01

    In the advanced neutron source reactor (ANSR) fuel element region, heat fluxes will be elevated. Early designs corresponded to average and estimated hot-spot fluxes of 11 to 12 and 21 to 22 MW/m 2 , respectively. Design changes under consideration may lower these values to ∼ 9 and 17 MW/m 1 . In either event, the development of a satisfactory burnout heat flux correlation is an important element among the many thermal-hydraulic design issues, since the critical power ratio will depend in part on its validity. Relatively little work in the area of subcooled-flow burnout has been published over the past 12 yr. The authors have compared seven burnout correlations and modifications therefore with several sets of experimental data, of which the most relevant to the ANS core are those referenced. The best overall agreement between the correlations tested and these data is currently provided by a modification of Thorgerson et al. correlation. The variable ranges of the experimental data are outlined and the results of the correlation comparisons are summarized

  7. Analysis and design of multilayer structures for neutron monochromators and supermirrors

    International Nuclear Information System (INIS)

    Masalovich, S.

    2013-01-01

    A relatively simple and accurate analytical model for studying the reflectivity of neutron multilayer monochromators and supermirrors is proposed. Design conditions that must be fulfilled in order to reach the maximum reflectivity are considered. The question of the narrowest bandwidth of a monochromator is discussed and the number of layers required to build such a monochromator is derived. Finally, we propose a new and efficient algorithm for synthesis of a supermirror with specified parameters and discuss some inherent restrictions on an attainable reflectivity. -- Highlights: • The inequality (not equation) that defines the thicknesses of layers was obtained. • Ready-to-use formula for the width of the spectral line was found. • Non-quarter-wave monochromators were suggested. • We propose a new algorithm for design of a neutron supermirror. • The problem of minimizing the number of layers in a supermirror is raised

  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. Electrostatic levitation facility optimized for neutron diffraction studies of high temperature liquids at a spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Mauro, N. A., E-mail: namauro@noctrl.edu [Department of Physics, North Central College, Naperville, Illinois 60540 (United States); Vogt, A. J. [Instrument and Source Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Derendorf, K. S. [Mechanical Engineering and Materials Science, Washington University, St. Louis, Missouri 63130 (United States); Johnson, M. L.; Kelton, K. F. [Department of Physics and Institute of Materials Science and Engineering, Washington University, 1 Brookings Drive, St. Louis, Missouri 63130 (United States); Rustan, G. E.; Quirinale, D. G.; Goldman, A. I. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Kreyssig, A. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Division of Materials Sciences and Engineering, Ames Laboratory, Ames, Iowa 50011 (United States); Lokshin, K. A. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Quantum Condensed Matter Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Neuefeind, J. C.; An, Ke [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Wang, Xun-Li [Department of Physics and Materials Science, City University of Hong Kong, 83 Tat Chee Ave., Kowloon (Hong Kong); Egami, T. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Department of Physics and Astronomy, Joint Institute for Neutron Sciences, University of Tennessee, Knoxville, Tennessee 37996 (United States)

    2016-01-15

    Neutron diffraction studies of metallic liquids provide valuable information about inherent topological and chemical ordering on multiple length scales as well as insight into dynamical processes at the level of a few atoms. However, there exist very few facilities in the world that allow such studies to be made of reactive metallic liquids in a containerless environment, and these are designed for use at reactor-based neutron sources. We present an electrostatic levitation facility, NESL (for Neutron ElectroStatic Levitator), which takes advantage of the enhanced capabilities and increased neutron flux available at spallation neutron sources (SNSs). NESL enables high quality elastic and inelastic neutron scattering experiments to be made of reactive metallic and other liquids in the equilibrium and supercooled temperature regime. The apparatus is comprised of a high vacuum chamber, external and internal neutron collimation optics, and a sample exchange mechanism that allows up to 30 samples to be processed between chamber openings. Two heating lasers allow excellent sample temperature homogeneity, even for samples approaching 500 mg, and an automated temperature control system allows isothermal measurements to be conducted for times approaching 2 h in the liquid state, with variations in the average sample temperature of less than 0.5%. To demonstrate the capabilities of the facility for elastic scattering studies of liquids, a high quality total structure factor for Zr{sub 64}Ni{sub 36} measured slightly above the liquidus temperature is presented from experiments conducted on the nanoscale-ordered materials diffractometer (NOMAD) beam line at the SNS after only 30 min of acquisition time for a small sample (∼100 mg)

  10. Study of a nTHGEM-based thermal neutron detector

    Science.gov (United States)

    Li, Ke; Zhou, Jian-Rong; Wang, Xiao-Dong; Xiong, Tao; Zhang, Ying; Xie, Yu-Guang; Zhou, Liang; Xu, Hong; Yang, Gui-An; Wang, Yan-Feng; Wang, Yan; Wu, Jin-Jie; Sun, Zhi-Jia; Hu, Bi-Tao

    2016-07-01

    With new generation neutron sources, traditional neutron detectors cannot satisfy the demands of the applications, especially under high flux. Furthermore, facing the global crisis in 3He gas supply, research on new types of neutron detector as an alternative to 3He is a research hotspot in the field of particle detection. GEM (Gaseous Electron Multiplier) neutron detectors have high counting rate, good spatial and time resolution, and could be one future direction of the development of neutron detectors. In this paper, the physical process of neutron detection is simulated with Geant4 code, studying the relations between thermal conversion efficiency, boron thickness and number of boron layers. Due to the special characteristics of neutron detection, we have developed a novel type of special ceramic nTHGEM (neutron THick GEM) for neutron detection. The performance of the nTHGEM working in different Ar/CO2 mixtures is presented, including measurements of the gain and the count rate plateau using a copper target X-ray source. A detector with a single nTHGEM has been tested for 2-D imaging using a 252Cf neutron source. The key parameters of the performance of the nTHGEM detector have been obtained, providing necessary experimental data as a reference for further research on this detector. Supported by National Natural Science Foundation of China (11127508, 11175199, 11205253, 11405191), Key Laboratory of Neutron Physics, CAEP (2013DB06, 2013BB04) and CAS (YZ201512)

  11. Design of auto-control high-voltage control system of pulsed neutron generator

    International Nuclear Information System (INIS)

    Lv Juntao

    2008-01-01

    It is difficult to produce multiple anode controlling time sequences under different logging mode for the high-voltage control system of the conventional pulsed neutron generator. It is also difficult realize sequential control among anode high-voltage, filament power supply and target voltage to make neutron yield stable. To these problems, an auto-control high-voltage system of neutron pulsed generator was designed. It not only can achieve anode high-voltage double blast time sequences, which can measure multiple neutron blast time sequences such as Σ, activated spectrum, etc. under inelastic scattering mode, but also can realize neutron generator real-time measurement of multi-state parameters and auto-control such as target voltage pulse width modulation (PWM), filament current, anode current, etc., there by it can produce stable neutron yield and realize stable and accurate measurement of the pulsed neutron full spectral loging tool. (authors)

  12. Detailed evaluation of ET-RR-2 neutronic design: approach to criticality

    International Nuclear Information System (INIS)

    Ashoub, N.; Amin, E.

    2000-01-01

    The ET-RR-2 safety analysis evaluation effort included the assessment of the reactor core nuclear design parameters. The present work complements the previous effort of neutronic calculations using NCNSRC computer code packages. This paper is the first in comprehensive neutronic calculation assessment up to the equilibrium core. It deals with evaluation of neutronic calculation submitted during the early phase of the licensing process; namely the proposal to assemble the first critical core, calculation of power peaking factor and determination of detailed energy group flux distribution in the core particularly in the cobalt irradiation position. The neutronic design criteria are examined qualitatively and quantitatively in the present paper

  13. Detailed Design of Cooling Water System for Cold Neutron Source in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Soo; Choi, Jung Woon; Kim, Y. K.; Wu, S. I.; Lee, Y. S

    2007-04-15

    To make cold neutron, a cryogenic refrigerator is necessary to transform moderator into cryogenic state so, thermal neutron is changed into cold neutron through heat transfer with moderator. A cryogenic refrigerator mainly consists of two apparatus, a helium compressor and a cold box which needs supply of cooling water. Therefore, cooling water system is essential to operate of cryogenic refrigerator normally. This report is mainly focused on the detailed design of the cooling water system for the HANARO cold neutron source, and describes design requirement, calculation, specification of equipment and water treatment method.

  14. Detailed Design of Cooling Water System for Cold Neutron Source in HANARO

    International Nuclear Information System (INIS)

    Kim, Bong Soo; Choi, Jung Woon; Kim, Y. K.; Wu, S. I.; Lee, Y. S.

    2007-04-01

    To make cold neutron, a cryogenic refrigerator is necessary to transform moderator into cryogenic state so, thermal neutron is changed into cold neutron through heat transfer with moderator. A cryogenic refrigerator mainly consists of two apparatus, a helium compressor and a cold box which needs supply of cooling water. Therefore, cooling water system is essential to operate of cryogenic refrigerator normally. This report is mainly focused on the detailed design of the cooling water system for the HANARO cold neutron source, and describes design requirement, calculation, specification of equipment and water treatment method

  15. Neutron capture studies of {sup 206}Pb at a cold neutron beam

    Energy Technology Data Exchange (ETDEWEB)

    Schillebeeckx, P.; Kopecky, S.; Quetel, C.R.; Tresl, I.; Wynants, R. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); Belgya, T.; Szentmiklosi, L. [Institute for Energy Security and Environmental Safety, Centre for Energy Research, Budapest (Hungary); Borella, A. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); SCK CEN, Mol (Belgium); Mengoni, A. [Nuclear Data Section, International Atomic Energy Agency (IAEA), Wagramerstrasse 5, PO Box 100, Vienna (Austria); Agenzia Nazionale per le Nuove Tecnologie, l' Energia e lo Sviluppo Economico Sostenibile (ENEA), Bologna (Italy)

    2013-11-15

    Gamma-ray transitions following neutron capture in {sup 206}Pb have been studied at the cold neutron beam facility of the Budapest Neutron Centre using a metallic sample enriched in {sup 206}Pb and a natural lead nitrate powder pellet. The measurements were performed using a coaxial HPGe detector with Compton suppression. The observed {gamma} -rays have been incorporated into a decay scheme for neutron capture in {sup 206}Pb. Partial capture cross sections for {sup 206}Pb(n, {gamma}) at thermal energy have been derived relative to the cross section for the 1884 keV transition after neutron capture in {sup 14}N. From the average crossing sum a total thermal neutron capture cross section of 29{sup +2}{sub -1} mb was derived for the {sup 206}Pb(n, {gamma}) reaction. The thermal neutron capture cross section for {sup 206}Pb has been compared with contributions due to both direct capture and distant unbound s-wave resonances. From the same measurements a thermal neutron-induced capture cross section of (649 {+-} 14) mb was determined for the {sup 207}Pb(n, {gamma}) reaction. (orig.)

  16. Design of calibration method in neutron and individual dosimeter

    International Nuclear Information System (INIS)

    Belkhodia, M.

    1984-12-01

    Usually albedo dosemeters are calibrated with beam of monoenergetic neutrons. Since neutron energy around neutron sources varies greatly, we applied the calibration method to a mixed field whose energy spectrum lies between 0.025 ev and 10 Mev. The method is based on a mathematical model that deals with the dosimeter response as a function at the neutron energy. The measurements carried out with solid state nuclear track detectors show the dosimeter practical aspect. The albedo dosimeter calibration gave results on good agreement with the international institution recommendations

  17. Neutronic study of the two french heavy water reactors

    International Nuclear Information System (INIS)

    Horowitz, J.

    1955-01-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

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

    International Nuclear Information System (INIS)

    Shahidan Radiman

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  20. Neutronic design of the RSG-GAS silicide core

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

    2002-07-01

    The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

  1. Study of liquid hydrogen and liquid deuterium cold neutron sources

    International Nuclear Information System (INIS)

    Harig, H.D.

    1969-01-01

    In view of the plant of the cold neutron source for a high flux reactor (maximal thermal flux of about 10 15 n/cm 2 s) an experimental study of several cold sources of liquid hydrogen and liquid deuterium has been made in a low power reactor (100 kW, about 10 12 n/cm 2 s). We have investigated: -cold neutron sources of liquid hydrogen shaped as annular layers of different thickness. Normal liquid hydrogen was used as well as hydrogen with a high para-percentage. -Cold neutron sources of liquid deuterium in cylinders of 18 and 38 cm diameter. In this case the sources could be placed into different positions to the reactor core within the heavy water reflector. This report gives a general description of the experimental device and deals more detailed with the design of the cryogenic systems. Then, the measured results are communicated, interpreted and finally compared with those of a theoretical study about the same cold moderators which have been the matter of the experimental investigation. (authors) [fr

  2. Design of collimator in the radial piercing beam port of Kartini reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    M Ilma Muslih A; Andang Widiharto; Yohannes Sardjono

    2014-01-01

    Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni (95%) as collimator wall, 15 cm thick of Al as moderator, 1 cm thick of Pb as γ-ray shielding, 1.5 cm thick of Boral as additional material, with 2 cm aperture diameter, epithermal neutron beam with maximum flux of 5.03 x 10 8 n.cm -2 .s -1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 2.17 x 10 -13 Gy.cm 2 .n -1 and 1.16 x 10 -13 Gy.cm 2 .n -l , minimum thermal neutron per epithermal neutron ratio of 0.12, and maximum directionality of 0.835 . It did not fully pass the IAEA's criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 10 9 n.cm -2 .s -l . Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 10 8 n.cm -2 .s -1 and fast neutron flux close to 2 x 10 -13 Gy.cm 2 .n -1 it is still feasible for BNCT in vivo experiment. (author)

  3. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  4. Accelerator conceptual design and needs of nuclear data for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Yamanaka, Toshiyuki; Yokobori, Hitoshi

    1999-01-01

    An optimization study has been made on an accelerator-based facility for the boron neutron capture therapy. The energy of the incident proton and the arrangement of the moderator assemblies are optimized. The beam current and the accelerating voltage are determined so that the accelerator power becomes minimum. The proposed facility is equipped with a 2.5 MeV proton accelerator of 10-25 mA, a lithium target, and a heavy water moderator contained in an aluminum tank. Each of these equipment is feasible, if proper R and D works have been done. Our new design requires the beam power of less than a hundred kW for the accelerator, although that of our previous design was 1 MW. The reduction of the beam power makes the cooling system for the target much simpler. The essential issues for realization of this concept are long-life lithium targets under high heat flux and high current proton accelerators with average currents of more than 10 mA. It is necessary for the reasonable design of a small-sized and low cost facility to get good accuracy nuclear reaction data. Especially, the latest Li/Be(p, n) neutron yield data in a range of threshold energy - few MeV are required for exact evaluation of neutron energy spectrum used therapy. And damage data by low energy proton beam are also important to evaluate integrity of target material. (author)

  5. Neutrons scattering studies in the actinide region

    International Nuclear Information System (INIS)

    Kegel, G.H.R.; Egan, J.J.

    1992-09-01

    During the report period were investigated the following areas: prompt fission neutron energy spectra measurements; neutron elastic and inelastic scattering from 239 Pu; neutron scattering in 181 Ta and 197 Au; response of a 235 U fission chamber near reaction thresholds; two-parameter data acquisition system; ''black'' neutron detector; investigation of neutron-induced defects in silicon dioxide; and multiple scattering corrections. Four Ph.D. dissertations and one M.S. thesis were completed during the report period. Publications consisted of three journal articles, four conference papers in proceedings, and eleven abstracts of presentations at scientific meetings. There are currently four Ph.D. and one M.S. candidates working on dissertations directly associated with the project. In addition, three other Ph.D. candidates are working on dissertations involving other aspects of neutron physics in this laboratory

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

    Science.gov (United States)

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

  7. A study of dissipative phenomena using Orion, a 4 π sectorized neutron detector

    International Nuclear Information System (INIS)

    Galin, J.; Guerreau, D.; Morjean, M.; Pouthas, J.; Saint-Laurent, F.; Sokolov, A.; Wang, X.M.; Piasecki, E.; Charvet, J.L.; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

    1990-01-01

    When studying the behavior of hot nuclei, the challenge is twofold: how are they formed in nucleus-nucleus collisions and how do they decay. For heavy and, thus neutron rich systems a large fraction of the thermalized energy is evacuated by neutron evaporation. Therefore the numbering, event-wise, of neutrons, over 4 π, gives a strong handle on energy dissipation for the different reaction channels. The first neutron measurements of this kind were performed using spherical detectors made of two hemispheres. Since then, a new and larger 4 π detector, ORION, has been designed in order to get information on the spatial distribution of the neutrons. The main characteristics of ORION are described and a few examples are given in order to illustrate the capabilities of such a detector in the study of dissipative collisions

  8. Feasibility study of a SiC sandwich neutron spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Jian, E-mail: caepwujian@163.com [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900, Sichuan Province (China); Lei, Jiarong, E-mail: jiarong_lei@163.com [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900, Sichuan Province (China); Jiang, Yong; Chen, Yu; Rong, Ru; Zou, Dehui; Fan, Xiaoqiang [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900, Sichuan Province (China); Chen, Gang; Li, Li; Bai, Song [Nanjing Electronic Devices Institute, Nanjing 210016 (China)

    2013-04-21

    Semiconductor sandwich neutron spectrometers are suitable for in-pile measurements of fast reactor spectra thanks to their compact and relatively simple design. We have assembled and tested a sandwich neutron spectrometer based on 4H-silicon carbide (4H-SiC) Schottky diodes. The SiC diodes detect neutrons via neutron-induced charged particles (tritons and alpha particles) produced by {sup 6}Li(n,α){sup 3}H reaction. {sup 6}LiF neutron converter layers are deposited on the front surface of Schottky diodes by magnetron sputtering. The responses of SiC diodes to charged particles were investigated with an {sup 241}Am alpha source. A sandwich neutron spectrometer was assembled with two SiC Schottky diodes selected based on the charged-particle-response experimental results. The low-energy neutron response of the sandwich spectrometer was measured in the neutron field of the Chinese Fast Burst Reactor-II (CFBR-II). Spectra of alpha particles and tritons from {sup 6}Li(n,α){sup 3}H reaction were obtained with two well-resolved peaks. The energy resolution of the sum spectrum was 8.8%. The primary experimental results confirmed the 4H-SiC sandwich neutron spectrometer's feasibility. -- Highlights: ► Sandwich neutron spectrometer employing 4H-SiC as a detecting material has been developed for the first time. ► {sup 6}LiF neutron converter has been deposited on the surface of 4H-SiC Schottky diode. ► Preliminary testing results obtained with the 4H-SiC sandwich neutron spectrometer are presented.

  9. Studies for improvement of WWER-440 neutron fluence determination

    International Nuclear Information System (INIS)

    Ilieva, Kr.; Belousov, S.; Apostolov, T.

    2001-01-01

    For assessment of radiation embrittlement and prediction of reactor vessel lifetime with reasonable conservatism a 'best estimated' neutron fluence is necessary. New studies purposed to improve the fluence determination are presented: 1) study on the reliability of multigroup presentation of the neutron cross sections, and 2) impact of negative gradient of reactor power in the periphery assemblies on the neutron fluence evaluation. The results of these studies are base for improvement of neutron fluence determination methodology applied by the INRNE, BAS at Kozloduy NPP. (author)

  10. Neutron streaming studies along JET shielding penetrations

    Science.gov (United States)

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

    2017-09-01

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

  11. Preliminary neutronic design of TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

    2001-01-01

    It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

  12. Design of neutron streak camera for fusion diagnostics

    International Nuclear Information System (INIS)

    Wang, C.L.; Kalibjian, R.; Singh, M.S.

    1982-06-01

    The D-T burn time for advanced laser-fusion targets is calculated to be very short, 2 . Each fission fragment leaving the cathode generates 400 secondary electrons that are all < 20 eV. These electrons are focussed to a point with an extractor and an anode, and are then purified with an electrostatic deflector. The electron beam is streaked and detected with the standard streak camera techniques. Careful shielding is needed for x-rays from the fusion target and general background. It appears that the neutron streak camera can be a viable and unique tool for studying temporal history of fusion burns in D-T plasmas of a few keV ion temperature

  13. A practical neutron shielding design based on data-base interpolation

    International Nuclear Information System (INIS)

    Jiang, S.H.; Sheu, R.J.

    1993-01-01

    Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

  14. Design and construction of a data acquisition system for a neutron diffractometer

    International Nuclear Information System (INIS)

    Baeza F, Lorena M.

    1995-01-01

    This work presents the design and construction of a data acquisition system for the neutron diffractometer of the La Reina Nuclear Studies Center. The system counts simultaneously the events produced in 17 detection channels, in 128 time channels with 10 μs duration, synchronized with a external signal. The event counting is performed by associating each detection and time channel to a binary code which determine a storage memory. The system is operated by a computer which controls the data acquisition and transfer. The designed software allows the data acquisition and storage in a file for later processing

  15. Optimization study of normal conductor tokamak for commercial neutron source

    Science.gov (United States)

    Fujita, T.; Sakai, R.; Okamoto, A.

    2017-05-01

    The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

  16. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  17. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  18. Neutron and gamma-ray toxicity studies

    International Nuclear Information System (INIS)

    Ainsworth, E.J.

    1975-01-01

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

  19. Study of neutron radiation effects on MOS structures

    International Nuclear Information System (INIS)

    Vaidya, Sangeeta J.

    2003-01-01

    We have studied charge trapping in the oxide and generation of interface states due to neutrons. It is observed that neutrons, though uncharged, are capable of causing ionization and interface damage and it is significant under biased irradiation conditions. One of the important features of this work is that neutron irradiation was carried out in a nuclear reactor (swimming pool type) itself in contrast to the earlier reported work which used separate neutron sources for similar studies. To simulate real life situations, all our devices were biased during irradiation. In our belief, both of these facts gave credence to our observed experimental results. (author)

  20. Progress in study of a medical reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Hirota, Jitsuya; Tamao, Shigeo; Kanda, Keiji; Mishima, Yutaka.

    1993-01-01

    A design study of a medical reactor for Boron Neutron Capture Therapy has made progress. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of hexagonal lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam holes are to be provided for simultaneous treatments by thermal and epithermal neutrons and for further biomedical research. The design objectives for the beam holes are to deliver the therapeutic doses in a modest time (30 to 60 min) with minimal fast neutron and gamma contaminants. The n-γ coupling Sn transport calculations have been carried out using n-21 and γ-9 group cross sections on 2-dim. practical models. The calculated results indicate that the design objectives will be achievable even if the thermal power of the reactor is reduced to 1 MW. (author)

  1. Applications of neutron scattering to the study of magnetic materials

    International Nuclear Information System (INIS)

    Koehler, W.C.

    1976-01-01

    The types of interactions that neutrons undergo with condensed matter are reviewed and those properties of neutrons that make them an ideal probe for the study of magnetism on a microscopic scale are discussed. Following a very brief survey of experimental methods, a few illustrative examples of specific investigations are described in sufficient detail to illustrate the power of the techniques. Views as to the future directions that may be taken by neutron scattering are presented

  2. Neutron diffraction studies of thin film multilayer structures

    International Nuclear Information System (INIS)

    Majkrzak, C.F.

    1985-01-01

    The application of neutron diffraction methods to the study of the microscopic chemical and magnetic structures of thin film multilayers is reviewed. Multilayer diffraction phenomena are described in general and in particular for the case in which one of the materials of a bilayer is ferromagnetic and the neutron beam polarized. Recent neutron diffraction measurements performed on some interesting multilayer systems are discussed. 70 refs., 5 figs

  3. Studies of neutron irradiation effects at IPNS-REF

    International Nuclear Information System (INIS)

    Kirk, M.A.

    1983-09-01

    Neutron irradiation effects studies at the Radiation Effects Facility (REF) at the Intense Pulsed Neutron Source (IPNS) located at Argonne National Laboratory (ANL) are reviewed. A brief history of the development of this user facility is followed by an overview of the scientific program. Experiments unique to a spallation neutron source are covered in more detail. Future direction of research at this facility is suggested

  4. Synthesis and study of neutron-rich nuclides

    International Nuclear Information System (INIS)

    Luo Yixiao

    1995-01-01

    During the past few years our understanding of the decay properties and nuclear structure has been extended in a systematic fashion for the neutron-rich nuclei. This review will first sketch the production and identification of the neutron-rich nuclei throughout the whole mass region, and will then discuss the impressive progress in the studies of the exotic decay properties and nuclear structure of neutron-rich nuclei. Their astrophysical implications will also be outlined

  5. Use of ultracold neutrons for condensed-matter studies

    Energy Technology Data Exchange (ETDEWEB)

    Michaudon, A.

    1997-05-01

    Ultracold neutrons have such low velocities that they are reflected by most materials at all incident angles and can be stored in material bottles for long periods of time during which their intrinsic properties can be studied in great detail. These features have been mainly used for fundamental-physics studies including the detection of a possible neutron electric dipole moment and the precise determination of neutron-decay properties. Ultracold neutrons can also play a role in condensed-matter studies with the help of high-resolution spectrometers that use gravity as a strongly dispersive medium for low-velocity neutrons. Such studies have so far been limited by the low intensity of existing ultracold-neutron sources but could be reconsidered with more intense sources, which are now envisaged. This report provides a broad survey of the properties of ultracold neutrons (including their reflectivity by different types of samples), of ultracold-neutron spectrometers that are compared with other high-resolution instruments, of results obtained in the field of condensed matter with these instruments, and of neutron microscopes. All these subjects are illustrated by numerous examples.

  6. Use of ultracold neutrons for condensed-matter studies

    International Nuclear Information System (INIS)

    Michaudon, A.

    1997-05-01

    Ultracold neutrons have such low velocities that they are reflected by most materials at all incident angles and can be stored in material bottles for long periods of time during which their intrinsic properties can be studied in great detail. These features have been mainly used for fundamental-physics studies including the detection of a possible neutron electric dipole moment and the precise determination of neutron-decay properties. Ultracold neutrons can also play a role in condensed-matter studies with the help of high-resolution spectrometers that use gravity as a strongly dispersive medium for low-velocity neutrons. Such studies have so far been limited by the low intensity of existing ultracold-neutron sources but could be reconsidered with more intense sources, which are now envisaged. This report provides a broad survey of the properties of ultracold neutrons (including their reflectivity by different types of samples), of ultracold-neutron spectrometers that are compared with other high-resolution instruments, of results obtained in the field of condensed matter with these instruments, and of neutron microscopes. All these subjects are illustrated by numerous examples

  7. 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

  8. National Array of Neutron Detectors (NAND): A versatile tool for nuclear reaction studies

    Energy Technology Data Exchange (ETDEWEB)

    Golda, K.S., E-mail: goldaks@gmail.com [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India); Jhingan, A.; Sugathan, P. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India); Singh, Hardev [Department of Physics, Kurukshetra University, Kurukshetra 136119 (India); Singh, R.P. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India); Behera, B.R. [Department of Physics, Panjab University, Chandigarh 160014 (India); Mandal, S. [Department of Physics and Astrophysics, Delhi University, New Delhi 110007 (India); Kothari, A.; Gupta, Arti; Zacharias, J.; Archunan, M.; Barua, P.; Venkataramanan, S.; Bhowmik, R.K. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India); Govil, I.M. [Department of Physics, Panjab University, Chandigarh 160014 (India); Datta, S.K.; Chatterjee, M.B. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India)

    2014-11-01

    The first phase of the National Array of Neutron Detectors (NAND) consisting of 26 neutron detectors has been commissioned at the Inter University Accelerator Centre (IUAC), New Delhi. The motivation behind setting up of such a detector system is the need for more accurate and efficient study of reaction mechanisms in the projectile energy range of 5–8 MeV/n using heavy ion beams from a 15 UD Pelletron and an upgraded LINAC booster facility at IUAC. The above detector array can be used for inclusive as well as exclusive measurements of reaction products of which at least one product is a neutron. While inclusive measurements can be made using only the neutron detectors along with the time of flight technique and a pulsed beam, exclusive measurements can be performed by detecting neutrons in coincidence with charged particles and/or fission fragments detected with ancillary detectors. The array can also be used for neutron tagged gamma-ray spectroscopy in (HI, xn) reactions by detecting gamma-rays in coincidence with the neutrons in a compact geometrical configuration. The various features and the performance of the different aspects of the array are described in the present paper. -- Highlights: •We report the design, fabrication and installation of a 26 element modular neutron detection system (NAND). •The array has been designed for the fusion–fission studies at near and above the barrier energies. •The relevant characteristics of the array are studied exhaustively and reported. •The efficiency of the detectors are measured and compared with the monte carlo simulations. •The second phase of the array will be augmented with 80 more neutron detectors which will enable the system to measure the neutron multiplicity distribution.

  9. National Array of Neutron Detectors (NAND): A versatile tool for nuclear reaction studies

    International Nuclear Information System (INIS)

    Golda, K.S.; Jhingan, A.; Sugathan, P.; Singh, Hardev; Singh, R.P.; Behera, B.R.; Mandal, S.; Kothari, A.; Gupta, Arti; Zacharias, J.; Archunan, M.; Barua, P.; Venkataramanan, S.; Bhowmik, R.K.; Govil, I.M.; Datta, S.K.; Chatterjee, M.B.

    2014-01-01

    The first phase of the National Array of Neutron Detectors (NAND) consisting of 26 neutron detectors has been commissioned at the Inter University Accelerator Centre (IUAC), New Delhi. The motivation behind setting up of such a detector system is the need for more accurate and efficient study of reaction mechanisms in the projectile energy range of 5–8 MeV/n using heavy ion beams from a 15 UD Pelletron and an upgraded LINAC booster facility at IUAC. The above detector array can be used for inclusive as well as exclusive measurements of reaction products of which at least one product is a neutron. While inclusive measurements can be made using only the neutron detectors along with the time of flight technique and a pulsed beam, exclusive measurements can be performed by detecting neutrons in coincidence with charged particles and/or fission fragments detected with ancillary detectors. The array can also be used for neutron tagged gamma-ray spectroscopy in (HI, xn) reactions by detecting gamma-rays in coincidence with the neutrons in a compact geometrical configuration. The various features and the performance of the different aspects of the array are described in the present paper. -- Highlights: •We report the design, fabrication and installation of a 26 element modular neutron detection system (NAND). •The array has been designed for the fusion–fission studies at near and above the barrier energies. •The relevant characteristics of the array are studied exhaustively and reported. •The efficiency of the detectors are measured and compared with the monte carlo simulations. •The second phase of the array will be augmented with 80 more neutron detectors which will enable the system to measure the neutron multiplicity distribution

  10. Radiochemical studies of neutron deficient actinide isotopes

    International Nuclear Information System (INIS)

    Williams, K.E.

    1978-04-01

    The production of neutron deficient actinide isotopes in heavy ion reactions was studied using alpha, gamma, x-ray, and spontaneous fission detection systems. A new isotope of berkelium, 242 Bk, was produced with a cross-section of approximately 10 μb in reactions of boron on uranium and nitrogen on thorium. It decays by electron capture with a half-life of 7.0 +- 1.3 minutes. The alpha-branching ratio for this isotope is less than 1% and the spontaneous fission ratio is less than 0.03%. Studies of (Heavy Ion, pxn) and (Heavy Ion, αxn) transfer reactions in comparison with (Heavy ion, xn) compound nucleus reactions revealed transfer reaction cross-sections equal to or greater than the compound nucleus yields. The data show that in some cases the yield of an isotope produced via a (H.I.,pxn) or (H.I.,αxn) reaction may be higher than its production via an xn compound nucleus reaction. These results have dire consequences for proponents of the ''Z 1 + Z 2 = Z/sub 1+2/'' philosophy. It is no longer acceptable to assume that (H.I.,pxn) and (H.I.,αxn) product yields are of no consequence when studying compound nucleus reactions. No evidence for spontaneous fission decay of 228 Pu, 230 Pu, 232 Cm, or 238 Cf was observed indicating that strictly empirical extrapolations of spontaneous fission half-life data is inadequate for predictions of half-lives for unknown neutron deficient actinide isotopes

  11. Radiochemical studies of neutron deficient actinide isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Williams, K.E.

    1978-04-01

    The production of neutron deficient actinide isotopes in heavy ion reactions was studied using alpha, gamma, x-ray, and spontaneous fission detection systems. A new isotope of berkelium, /sup 242/Bk, was produced with a cross-section of approximately 10 ..mu..b in reactions of boron on uranium and nitrogen on thorium. It decays by electron capture with a half-life of 7.0 +- 1.3 minutes. The alpha-branching ratio for this isotope is less than 1% and the spontaneous fission ratio is less than 0.03%. Studies of (Heavy Ion, pxn) and (Heavy Ion, ..cap alpha..xn) transfer reactions in comparison with (Heavy ion, xn) compound nucleus reactions revealed transfer reaction cross-sections equal to or greater than the compound nucleus yields. The data show that in some cases the yield of an isotope produced via a (H.I.,pxn) or (H.I.,..cap alpha..xn) reaction may be higher than its production via an xn compound nucleus reaction. These results have dire consequences for proponents of the ''Z/sub 1/ + Z/sub 2/ = Z/sub 1+2/'' philosophy. It is no longer acceptable to assume that (H.I.,pxn) and (H.I.,..cap alpha..xn) product yields are of no consequence when studying compound nucleus reactions. No evidence for spontaneous fission decay of /sup 228/Pu, /sup 230/Pu, /sup 232/Cm, or /sup 238/Cf was observed indicating that strictly empirical extrapolations of spontaneous fission half-life data is inadequate for predictions of half-lives for unknown neutron deficient actinide isotopes.

  12. Simulation study on the cold neutron guides in China advanced research reactor

    International Nuclear Information System (INIS)

    Guo Liping; Yang Tonghua; Wang Hongli; Sun Kai; Zhao Zhixiang

    2003-01-01

    The designs of the two cold neutron guides, CNG1 and CNG2, to be built in China advanced research reactor (CARR) are studied with Monte-Carlo simulation technique. The neutron flux density at the exit of the both guides can reach above 1 x10 9 cm -2 ·s -1 under the assumed flux spectrum of the cold neutron source. The transmission efficiency is 50% and 42%, and the maximum divergence is about 2.2 degree and 1.9 degree, respectively for CNG1 and CNG2. Neutron distribution along horizontal direction is quite uniform for both guides, with maximum fluctuation of less than 3%. Gravity can affect neutron distribution along vertical direction considerably

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  14. A review on the study of polymer properties by Small Angle Neutron Scattering

    International Nuclear Information System (INIS)

    Seong, Baek Seok; Lee, Chang Hee; Sim, Hae Seop; Lee, Jung Sool; Kim, Hong Doo; Kim, Eu Gene; Cha, Kuk Heon

    1998-05-01

    This report contains concept of small angle neutron scattering , various design features and considerations of the small angle neutron spectrometer at HANARO, and recent trends of polymer studies by using this SANS technique with the installation of the spectrometer in near future. We, therefore, wish to review feasibility of small angle studies for polymer field at this spectrometer and to help possible beam time users for their experimental consideration. (author). 23 refs., 7 tabs., 23 figs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  17. A neutron scattering study of DCN

    International Nuclear Information System (INIS)

    Mackenzie, G.A.; Pawley, G.S.

    1979-01-01

    Phonons in deuterium cyanide have been measured by neutron coherent inelastic scattering. The main subject of study was the transverse acoustic mode in the (110) direction polarised along (110) which is associated with the first-order structural phase transition at 160 K. Measurements have shown that the frequency decreases by about 25% between about 225 and 160 K as the transition temperature is approached. The other acoustic modes observable in the a*b* scattering plane have been measured and show no anomalous temperature dependence. Optic modes were unobservable because of the small size of the single-crystal sample which gave insufficient scattered intensity. Apart from the 'soft' mode, the measured frequencies are in good agreement with lattice dynamics calculations. (author)

  18. Polarized Epithermal Neutron Studies of Magnetic Domains

    International Nuclear Information System (INIS)

    Alfimenkov, V.P.; Chernikov, A.N.; Lason, L.; Mareev, Yu. D.; Novitsky, V.V.; Pikelner, L.B.; Skoy, V.R.; Tsulaya, M.I.; Gould, C.R.; Haase, D.G.; Roberson, N.R.

    1997-01-01

    The average size and shape of magnetic domains in a material can be determined from the precession of polarized neutrons traversing the material. Epithermal neutrons (0.5eV< En<100eV), which process more slowly than thermals, effectively probe the internal structure of samples that are thick or have large domains or large internal fields. Such epithermal neutron measurements require a neutron polarizer and analyzer based on cryogenically polarized spin filters. We discuss the measurements at JINR, Dubna, of magnetic domains in a 2.0 cm. diam. crystal of holmium using 1.7 to 59eV neutrons polarized by a dynamically polarized proton target and analyzed with a statically polarized dysprosium target

  19. Polarized epithermal neutron studies of magnetic domains

    International Nuclear Information System (INIS)

    Alfimenkov, V.P.; Chernikov, A.N.; Lason, L.; Mareev, Y.D.; Novitsky, V.V.; Pikelner, L.B.; Skoy, V.R.; Tsulaya, M.I.; Gould, C.R.; Haase, D.G.; the Triangle Universities Nuclear Laboratory, Durham, North Carolina; Roberson, N.R.; the Triangle Universities Nuclear Laboratory, Durham, North Carolina

    1997-01-01

    The average size and shape of magnetic domains in a material can be determined from the precession of polarized neutrons traversing the material. Epithermal neutrons (0.5eV n <100eV), which precess more slowly than thermals, effectively probe the internal structure of samples that are thick or have large domains or large internal fields. Such epithermal neutron measurements require a neutron polarizer and analyzer based on cryogenically polarized spin filters. We discuss the measurement at JINR, Dubna, of magnetic domains in a 2.0 cm. diam. crystal of holmium using 1.7 to 59 eV neutrons polarized by a dynamically polarized proton target and analyzed with a statically polarized dysprosium target. copyright 1997 American Institute of Physics

  20. Technical design report of spallation neutron source facility in J-PARC

    International Nuclear Information System (INIS)

    Sakamoto, Shinichi

    2012-02-01

    One of the experimental facilities in Japan Proton Accelerator Research Complex (J-PARC) is the Materials and Life Science Experimental Facility (MLF), where high-intensity neutron beams are used as powerful probes for basic research on materials and life science, as well as research and development in industrial engineering. Neutrons are generated with nuclear spallation reaction by bombarding a mercury target with high-intensity proton beams. The neutrons are slowed down with supercritical hydrogen moderators and then extracted as beams to each experimental apparatus. The principal design of the spallation neutron source is compiled in this comprehensive report. (author)

  1. Design of a setup for {sup 252}Cf neutron source for storage and analysis purpose

    Energy Technology Data Exchange (ETDEWEB)

    Hei, Daqian [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Zhuang, Haocheng [Xi’an Middle School of Shanxi Province, Xi’an 710000 (China); Jia, Wenbao, E-mail: jiawenbao@163.com [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Collaborative Innovation Center of Radiation Medicine of Jiangsu Higher Education Institutions, Suzhou 215000 (China); Cheng, Can; Jiang, Zhou; Wang, Hongtao [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Chen, Da [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Collaborative Innovation Center of Radiation Medicine of Jiangsu Higher Education Institutions, Suzhou 215000 (China)

    2016-11-01

    {sup 252}Cf is a reliable isotopic neutron source and widely used in the prompt gamma ray neutron activation analysis (PGNAA) technique. A cylindrical barrel made by polymethyl methacrylate contained with the boric acid solution was designed for storage and application of a 5 μg {sup 252}Cf neutron source. The size of the setup was optimized with Monte Carlo code. The experiments were performed and the results showed the doses were reduced with the setup and less than the allowable limit. The intensity and collimating radius of the neutron beam could also be adjusted through different collimator.

  2. Proceedings of a workshop on methods for neutron scattering instrumentation design

    International Nuclear Information System (INIS)

    Hjelm, R.P.

    1997-09-01

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop. The international gathering of about 50 participants representing 15 national facilities, universities and corporations featured oral presentations, posters, discussions and demonstrations. Participants looked at a number of issues concerning neutron scattering instruments and the tools used in instrument design. Objectives included: (1) determining the needs of the neutron scattering community in instrument design computer code and information sharing to aid future instrument development, (2) providing for a means of training scientists in neutron scattering and neutron instrument techniques, and (3) facilitating the involvement of other scientists in determining the characteristics of new instruments that meet future scientific objectives, and (4) fostering international cooperation in meeting these needs. The scope of the meeting included: (1) a review of x-ray scattering instrument design tools, (2) a look at the present status of neutron scattering instrument design tools and models of neutron optical elements, and (3) discussions of the present and future needs of the neutron scattering community. Selected papers were abstracted separately for inclusion to the Energy Science and Technology Database

  3. A Workshop on Methods for Neutron Scattering Instrument Design. Introduction and Summary

    International Nuclear Information System (INIS)

    Hjelm, Rex P.

    1996-09-01

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop on ''Methods for Neutron Scattering Instrument Design'' September 23-25 at the E.O. Lawrence Berkeley National Laboratory. These proceedings are a collection of a portion of the invited and contributed presentations

  4. Proceedings of a workshop on methods for neutron scattering instrumentation design

    Energy Technology Data Exchange (ETDEWEB)

    Hjelm, R.P. [ed.] [Los Alamos National Lab., NM (United States)

    1997-09-01

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop. The international gathering of about 50 participants representing 15 national facilities, universities and corporations featured oral presentations, posters, discussions and demonstrations. Participants looked at a number of issues concerning neutron scattering instruments and the tools used in instrument design. Objectives included: (1) determining the needs of the neutron scattering community in instrument design computer code and information sharing to aid future instrument development, (2) providing for a means of training scientists in neutron scattering and neutron instrument techniques, and (3) facilitating the involvement of other scientists in determining the characteristics of new instruments that meet future scientific objectives, and (4) fostering international cooperation in meeting these needs. The scope of the meeting included: (1) a review of x-ray scattering instrument design tools, (2) a look at the present status of neutron scattering instrument design tools and models of neutron optical elements, and (3) discussions of the present and future needs of the neutron scattering community. Selected papers were abstracted separately for inclusion to the Energy Science and Technology Database.

  5. Designing research of fast neutron radiation field based on the reactor

    International Nuclear Information System (INIS)

    Zhang Wenzhong; Zhang Xiaomin

    2009-01-01

    Based on the Tsinghua University experimental nuclear reactor neutron source, this research designed moderate theory technical scheme, and the thickness of materials in the scheme were selected by means of Monte Carlo simulating method. An fast neutron radiation field was gained. (authors)

  6. Considerations in the design of an improved transportable neutron spectrometer

    CERN Document Server

    Williams, A M; Brushwood, J M; Beeley, P A

    2002-01-01

    The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.

  7. The low power miniature neutron source reactors: Design, safety and applications

    International Nuclear Information System (INIS)

    Ahmed, Y.A.; Ewa, I.O.B.; Umar, M.; Bezboruah, T.; Johri, M.; Akaho, E.H.K.

    2006-04-01

    The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 10 12 n.cm -2 .s -1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

  8. Experimental study on a cold neutron source of solid methylbenzene

    Energy Technology Data Exchange (ETDEWEB)

    Utsuro, M; Sugimoto, M; Fujita, Y [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1975-10-01

    An experimental study to produce cold neutrons with low temperature solid mesitylene as cold moderator in liquid helium and liquid nitrogen cryostats is reported. Measured cold neutron spectra by using an electron linac and time-of-flight method shows that this material is a better cold moderator than light water ice, giving the cold neutron output not so much inferior to that of solid methane in the temperature range above about 20 K and in the neutron energy region above about 1 MeV.

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

    Science.gov (United States)

    Ghassoun, J; Mostacci, D

    2011-08-01

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

  10. Basic design of the HANARO cold neutron source using MCNP code

    International Nuclear Information System (INIS)

    Yu, Yeong Jin; Lee, Kye Hong; Kim, Young Jin; Hwang, Dong Gil

    2005-01-01

    The design of the Cold Neutron Source (CNS) for the HANARO research reactor is on progress. The CNS produces neutrons in the low energy range less than 5meV using liquid hydrogen at around 21.6 K as the moderator. The primary goal for the CNS design is to maximize the cold neutron flux with wavelengths of around 2 ∼ 12 A and to minimize the nuclear heat load. In this paper, the basic design of the HANARO CNS is described

  11. Experimental study on the optimization for neutron emission in a small fast plasma focus operated at tens of Joules

    International Nuclear Information System (INIS)

    Tarifeno-Saldivia, A; Soto, L

    2014-01-01

    This work reports results of a systematic experimental study dealing with the optimization for neutron emission of the PF-50J plasma focus. The device was operated in a repetitive mode at repetition rates of 0.1-0.5 Hz. Optimal configurations, neutron emission rates, observed anisotropy, analysis and contrast of 'good' and 'bad' shots are currently presented. Additionally, engineering aspects on the neutron emission were also studied, such as contaminants removal circuit and chamber design.

  12. Overview of the Conceptual Design of the Future VENUS Neutron Imaging Beam Line at the Spallation Neutron Source

    Science.gov (United States)

    Bilheux, Hassina; Herwig, Ken; Keener, Scott; Davis, Larry

    VENUS (Versatile Neutron Imaging Beam line at the Spallation Neutron Source) will be a world-class neutron-imaging instrument that will uniquely utilize the Spallation Neutron Source (SNS) time-of-flight (TOF) capabilities to measure and characterize objects across several length scales (mm to μm). When completed, VENUS will provide academia, industry and government laboratories with the opportunity to advance scientific research in areas such as energy, materials, additive manufacturing, geosciences, transportation, engineering, plant physiology, biology, etc. It is anticipated that a good portion of the VENUS user community will have a strong engineering/industrial research focus. Installed at Beam line 10 (BL10), VENUS will be a 25-m neutron imaging facility with the capability to fully illuminate (i.e., umbra illumination) a 20 cm x 20 cm detector area. The design allows for a 28 cm x 28 cm field of view when using the penumbra to 80% of the full illumination flux. A sample position at 20 m will be implemented for magnification measurements. The optical components are comprised of a series of selected apertures, T0 and bandwidth choppers, beam scrapers, a fast shutter to limit sample activation, and flight tubes filled with Helium. Techniques such as energy selective, Bragg edge and epithermal imaging will be available at VENUS.

  13. Preliminary design of the Neutron Spectral Shifter that is dedicated to the IFMIF Liquid Breeder Validation Module

    Energy Technology Data Exchange (ETDEWEB)

    Mas, A., E-mail: amassanchez@gmail.com; Mota, F.; Casal, N.; García, A.; Rapisarda, D.; Arroyo, J.M.; Molla, J.; Ibarra, A.

    2014-10-15

    The International Fusion Materials Irradiation Facility (IFMIF) has a D-Li neutron stripping source that provides typical fusion irradiation conditions for material testing. The Liquid Breeder Validation Module (LBVM) is one of the medium flux test modules of the IFMIF that is used to account for some of the DEMO liquid breeder blanket R and D needs. Previous analyses have shown that the main irradiation parameters (He (appm)/dpa and H (appm)/dpa) in the medium flux area of the IFMIF can be improved to fit the expected parameters in the DEMO reactor for functional materials of liquid breeder blankets. Therefore, the design of an additional module, called the Neutron Spectral Shifter (NSS), has been considered to optimize the irradiation conditions of LBVM experiments. The proposed concept consists of supported tungsten plates working as a shifter material inside a steel structure. This design assures the mechanical integrity of the different components and it fulfills the neutronic requirements as well as the cooling capability. This present paper summarizes the work devoted to the design of the LBVM Neutron Spectral Shifter as well as the results of neutronic, thermo-hydraulic, mechanical and safety studies carried out to validate the design.

  14. Study on Neutron Generation by Using Modified Prototype D-D Neutron Generator

    International Nuclear Information System (INIS)

    Kim, In-Jung; Kim, Suk-Kwon; Park, Chang-Su; Jung, Nam-Suk; Jung, Hwa-Dong; Chung, Kyoung-Jae; Hwang, Yong-Seok; Choi, H. D.

    2006-01-01

    The effects of Ti target thickness and deuteron beam energy on neutron generation in the modified prototype DD neutron generator were studied. Three kinds of Ti targets with the thickness of 10 μm, 40 μm and 1 mm were used. Deuteron beam energy was varied from 45 keV to 65 keV. The effects of target thickness and deuteron beam energy were evaluated for every set of experimental run and the results were discussed

  15. Studies of neutron emission from relativistic nuclear interactions

    CERN Document Server

    Guo, S L; Wang, Y L; Guo, H Y; Sá Ben-Hao; Zheng, Y M; Brandt, R; Vater, P; Wan, J S; Ochs, M; Kulakov, B A; Sosnin, A N; Krivopustov, M I; Butsev, V S; Bradnova, V

    1999-01-01

    Studies were carried out on the yields and spatial distributions of secondary neutrons produced in the relativistic nuclear interactions of 1.5 GeV to 14.4 GeV projectiles p, d and alpha-particles with targets Pb and U/Pb. CR-39 track detectors were used to measure the neutrons. It shows that: (1) Secondary neutrons are produced in the whole length of Pb or U targets having a thickness of 20 cm. The neutron intensities produced by proton bombardments are reduced along the proton beam direction in the targets. The higher the energy of protons, the lower the reduction rate of the neutrons. The reduction rate of neutrons in U target is higher than in Pb target for the same energy of protons. (2) The radial intensities of neutrons decrease as the distance increases from the target central line. (3) The neutron yield in U target by proton bombardments is approx 55% higher than in Pb target. (4) The ratio of neutron yield by 14.4 GeV alpha to 7.3 GeV d bombardment in Pb target is 1.74+-0.20.

  16. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    International Nuclear Information System (INIS)

    Williams, P.T.; Lucas, A.T.

    1995-08-01

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  17. The neutron long counter NERO for studies of β-delayed neutron emission in the r-process

    International Nuclear Information System (INIS)

    Pereira, J.; Hosmer, P.; Lorusso, G.; Santi, P.; Couture, A.; Daly, J.; Del Santo, M.; Elliot, T.

    2010-01-01

    The neutron long counter NERO was built at the National Superconducting Cyclotron Laboratory (NSCL), Michigan State University, for measuring β-delayed neutron-emission probabilities. The detector was designed to work in conjunction with a β-delay implantation station, so that β decays and β-delayed neutrons emitted from implanted nuclei can be measured simultaneously. The high efficiency of about 40%, for the range of energies of interest, along with the small background, are crucial for measuring β-delayed neutron emission branchings for neutron-rich r-process nuclei produced as low intensity fragmentation beams in in-flight separator facilities.

  18. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

  19. IB: A Monte Carlo simulation tool for neutron scattering instrument design under PVM and MPI

    International Nuclear Information System (INIS)

    Zhao Jinkui

    2011-01-01

    Design of modern neutron scattering instruments relies heavily on Monte Carlo simulation tools for optimization. IB is one such tool written in C++ and implemented under Parallel Virtual Machine and the Message Passing Interface. The program was initially written for the design and optimization of the EQ-SANS instrument at the Spallation Neutron Source. One of its features is the ability to group simple instrument components into more complex ones at the user input level, e.g. grouping neutron mirrors into neutron guides and curved benders. The simulation engine manages the grouped components such that neutrons entering a group are properly operated upon by all components, multiple times if needed, before exiting the group. Thus, only a few basic optical modules are needed at the programming level. For simulations that require higher computer speeds, the program can be compiled and run in parallel modes using either the PVM or the MPI architectures.

  20. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Directory of Open Access Journals (Sweden)

    Ersez Tunay

    2017-01-01

    Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  1. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Science.gov (United States)

    Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

    2017-09-01

    The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  2. Some neutronic studies on flux-trap type moderator

    International Nuclear Information System (INIS)

    Kiyanagi, Y.; Watanabe, N.

    1991-01-01

    The neutronic performance of a flux-trap type moderator was studied by computer simulation in connection with the KENS-II target-moderator system. It was confirmed that this system can provide 1.3-1.4 times higher neutron intensity than a traditional wing-geometry moderator system. (author)

  3. Studies of neutron dissociation at Fermilab energies

    International Nuclear Information System (INIS)

    Ferbel, T.

    1975-01-01

    The latest results obtained in a continuing investigation of neutron dissociation in (pπ - ) systems in neutron--nucleus collisions between 50 and 300 GeV/c are summarized. The nuclear coherent dissociation data are discussed first; then new measurements of total cross sections of neutrons on nuclei in the Fermilab momentum range are presented; finally, neutron dissociation using a hydrogen target is considered, and the hydrogen data are compared with expectations from simple Deck models. A substantial correlation was observed between the mass and the t of the system produced. The spin structure of the pπ - amplitudes at low mass was described surprisingly well by the simple Deck mechanism. The t-channel helicity amplitudes contained comparable contributions from flip and nonflip terms, and the states produced were not restricted to those expected on the basis of the Morrison rule. (19 figures, 2 tables) (U.S.)

  4. System design considerations for fast-neutron interrogation systems

    International Nuclear Information System (INIS)

    Micklich, B.J.; Curry, B.P.; Fink, C.L.; Smith, D.L.; Yule, T.J.

    1993-01-01

    Nonintrusive interrogation techniques that employ fast neutrons are of interest because of their sensitivity to light elements such as carbon, nitrogen, and oxygen. The primary requirement of a fast-neutron inspection system is to determine the value of atomic densities, or their ratios, over a volumetric grid superimposed on the object being interrogated. There are a wide variety of fast-neutron techniques that can provide this information. The differences between the various nuclear systems can be considered in light of the trade-offs relative to the performance requirements for each system's components. Given a set of performance criteria, the operational requirements of the proposed nuclear systems may also differ. For instance, resolution standards will drive scanning times and tomographic requirements, both of which vary for the different approaches. We are modelling a number of the fast-neutron interrogation techniques currently under consideration, to include Fast Neutron Transmission Spectroscopy (FNTS), Pulsed Fast Neutron Analysis (PFNA), and its variant, 14-MeV Associated Particle Imaging (API). The goals of this effort are to determine the component requirements for each technique, identify trade-offs that system performance standards impose upon those component requirements, and assess the relative advantages and disadvantages of the different approaches. In determining the component requirements, we will consider how they are driven by system performance standards, such as image resolution, scanning time, and statistical uncertainty. In considering the trade-offs between system components, we concentrate primarily on those which are common to all approaches, for example: source characteristics versus detector array requirements. We will then use the analysis to propose some figures-of-merit that enable performance comparisons between the various fast-neutron systems under consideration. The status of this ongoing effort is presented

  5. Progress in neutron beam development at the HFR Petten (feasibility study for a BNCT facility)

    International Nuclear Information System (INIS)

    Constantine, G.; Moss, R.L.; Watkins, P.R.D.; Perks, C.A.; Delafield, H.J.; Ross, D.; Voorbraak, W.P.; Paardekooper, A.; Freudenreich, W.E.; Stecher-Rasmussen, F.

    1990-08-01

    Boron Neutron Capture Therapy, using intermediate energy neutrons to achieve the deep penetration essential for treating brain tumours, can be implemented with a filtered reactor neutron beam. This is designed to minimize the mean energy of the neutrons to keep proton recoil damage to the scalp within normal tissue tolerance limits whilst delivering the required thermal neutron fluence to the tumour over a reasonably short period. This can only be realized in conjunction with a high power density reactor. At the Joint Research Centre Petten an optimized neutron filter is currently being built for installation into the HB11 beam tube of the High Flux Reactor HFR. Part of the development leading to this design has been an extensive study of broad spectrum, filtered beam performance on the HB7 beam tube facility. A wide range of calculations was performed using the Monte Carlo code, MCPN, supported by validation experiments in which several filter configuration incorporating aluminium, sulphur, liquid argon, titanium and cadmium were installed for low power measurements of the neutron fluence rate, neutron spectra and beam gamma-ray contamination. The measurements were carried out within a successful European collaboration. Evaluations were made of the reactor core edge and unfiltered beam spectra, for comparison with MCNP calculations. Multi-foil activation methods and also gamma dose determination in the filtered beam using thermo-luminescent detectors were performed by the ECN. The Harwell/ Birmingham University collaborators undertook the neutron spectrum measurements in the filtered beam. proton recoil spectrometry was used above 30 keV, combined with a multi-sphere and BF 3 chamber response modification technique. Subsequent spectrum adjustment was carried out with the SENSAK code. The agreement between the calculated and measured spectra has given confidence in the reactor and filter modelling methods used to design the HB11 therapy facility. (author). 12 refs

  6. Nomenclature and principle of neutron humidistats design and methods of their checking

    International Nuclear Information System (INIS)

    Chaladze, A.P.; Melkumyan, V.E.

    1980-01-01

    The state of neutron hydrometry in ferrous metallurgy is considered. The nomenclature and technical characteristics of neutron humidistats and methods of their testing are presented as well as the local testing diagram for imitator certification and the testing of devices. Taking into account the design, neutron humidistats can be classified into two- and three-channel. As regards their structural realization, humidistats are classified into devices of the external type designed for measuring humidity in technological capacities and devices of the superposition type, designed for measuring the humidity of the material on a moving conveyer. The design of imitators for all types of humidistats is similar, that is the use of neutron retarders and absorbers, displaced relatively to each other [ru

  7. Design of a novel instrument for active neutron interrogation of artillery shells.

    Science.gov (United States)

    Bélanger-Champagne, Camille; Vainionpää, Hannes; Peura, Pauli; Toivonen, Harri; Eerola, Paula; Dendooven, Peter

    2017-01-01

    The most common explosives can be uniquely identified by measuring the elemental H/N ratio with a precision better than 10%. Monte Carlo simulations were used to design two variants of a new prompt gamma neutron activation instrument that can achieve this precision. The instrument features an intense pulsed neutron generator with precise timing. Measuring the hydrogen peak from the target explosive is especially challenging because the instrument itself contains hydrogen, which is needed for neutron moderation and shielding. By iterative design optimization, the fraction of the hydrogen peak counts coming from the explosive under interrogation increased from [Formula: see text]% to [Formula: see text]% (statistical only) for the benchmark design. In the optimized design variants, the hydrogen signal from a high-explosive shell can be measured to a statistics-only precision better than 1% in less than 30 minutes for an average neutron production yield of 109 n/s.

  8. Design of a novel instrument for active neutron interrogation of artillery shells.

    Directory of Open Access Journals (Sweden)

    Camille Bélanger-Champagne

    Full Text Available The most common explosives can be uniquely identified by measuring the elemental H/N ratio with a precision better than 10%. Monte Carlo simulations were used to design two variants of a new prompt gamma neutron activation instrument that can achieve this precision. The instrument features an intense pulsed neutron generator with precise timing. Measuring the hydrogen peak from the target explosive is especially challenging because the instrument itself contains hydrogen, which is needed for neutron moderation and shielding. By iterative design optimization, the fraction of the hydrogen peak counts coming from the explosive under interrogation increased from [Formula: see text]% to [Formula: see text]% (statistical only for the benchmark design. In the optimized design variants, the hydrogen signal from a high-explosive shell can be measured to a statistics-only precision better than 1% in less than 30 minutes for an average neutron production yield of 109 n/s.

  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. Design of a new neutron delivery system for high flux source

    International Nuclear Information System (INIS)

    Boffy, Romain

    2016-01-01

    these materials has been ascribed to this reaction. The methodology employed by this study consisted in understanding the general structure of borosilicates and how they behave under irradiation. Such materials have a microscopic structure strongly dependent upon their chemical content and particularly on the ratios between network formers and modifiers. The materials have been characterized by a suite of macroscopic and structural techniques such as hardness, TEM, Raman, SANS, etc, and their behaviour under irradiation was analysed. Macroscopic properties related to their resistance when used as guide structural elements were monitored. Also, changes in the vitreous structure due to radiation were observed by means of several experimental tools. For this purpose, an irradiation apparatus has been designed and manufactured to enable irradiation with thermal neutrons within the ILL reactor while keeping the samples below 100 C. The main advantage of this equipment, compared to others previously available, was that it enabled the glass to reach an equivalent neutron dose to that accumulated after several years of use as guides, in just few days. The concurrent use of complementary characterization techniques lead to the discovery that the studied glasses were deeply different in terms of their glass network. This had a strong impact on their macroscopic properties and their behaviour under irradiation. This result was a surprise since, as stated above, some of these materials were well known by the neutron guide manufacturers, and were considered to be almost equivalent because of their similar boron oxide content. The N-BK7 and S-BSL7 materials appear to be fairly homogeneous glasses at different length scales. More specifically, at nanometer scales, silicon and boron oxide units seem to mix and generate larger structures somewhat resembling crystalline Reedmergnerite. In contrast, N-ZK7 and Borofloat are characterized by either silicon or boron rich domains. As one

  11. Neutronics Study of the KANUTER Space Propulsion Reactor

    International Nuclear Information System (INIS)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee

    2014-01-01

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe 135 , and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity

  12. Neutronics Study of the KANUTER Space Propulsion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe{sup 135}, and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity.

  13. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2006-12-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (author)

  14. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2008-01-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (authors)

  15. Physics study of D-D/D-T neutron driven experimental subcritical assembly

    International Nuclear Information System (INIS)

    Sinha, Amar

    2015-01-01

    An experimental program to design and study external source driven subcritical assembly has been initiated at BARC. This program is aimed at understanding neutronic characteristics of accelerator driven system at low power level. In this series, a zero-power, sub-critical assembly driven by a D-D/D-T neutron generator has been developed. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The subcritical core is coupled to Purnima Neutron Generator. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k s and external neutron source efficiency φ* in great details. Some experiments with D-D and D-T neutrons have been presented. (author)

  16. Project study of a small-angle neutron scattering apparatus

    International Nuclear Information System (INIS)

    Schedler, E.; Pollet, J.L.

    1979-03-01

    This design study deals with the set up of a low angle scattering apparatus in the HMI reactor hall in Berlin. The experiences of other institutes with facilities of a similar type, - especially with D11 and D17 of the ILL in Grenoble, the set up the KFA in Juelich and of the PTB in Braunschweig -, are included to a large extend. The aim of this paper is - to define the necessary boundary conditions for the construction (including: installation of a cold source, the beam line, the neutron guide pipe and an extention of the reactor hall), -to determine the properties of the planned apparatus, especially in comparison with D11, probably the most versatile instrument, - to make desitions for the design of the components, - to work out the detailed drawings for construction - to estimate the costs and the time necessary for construction, if industrial manufacturers set up the project. (orig.) [de

  17. Neutron and gamma-ray toxicity studies

    International Nuclear Information System (INIS)

    Ainsworth, E.J.

    1975-01-01

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

  18. Computational Benchmark Calculations Relevant to the Neutronic Design of the Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.

    1999-01-01

    The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements

  19. Cylindrical IEC neutron source design for driven research reactor operation

    International Nuclear Information System (INIS)

    Miley, G.H.; Ulmen, B.; Amadio, G.; Leon, H.; Hora, H.

    2009-01-01

    A resurgence in nuclear power use is now underway worldwide. However, due many university research reactors shutdown, they must rely on using subcritical assemblies which employs a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The source is inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory. (author)

  20. Neutronics design for lead-bismuth cooled accelerator-driven system for transmutation of minor actinide

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi; Sasa, Toshinobu; Nishihara, Kenji; Oigawa, Hiroyuki; Takano, Hideki

    2004-01-01

    Neutronics design study was performed for lead-bismuth cooled accelerator-driven system (ADS) to transmute minor actinides. Early study for ADS indicated two problems: a large burnup reactivity swing and a significant peaking factor. To solve these problems, effect of design parameters on neutronics characteristics were searched. The design parameters were initial plutonium loading, buffer region between spallation target and core, and zone fuel loading. Parametric survey calculations were performed considering fuel cycle consisting of burnup and recycle. The results showed that burnup reactivity swing depends on the plutonium fraction in the initial fuel loading, and the lead-bismuth buffer region and the two-zone loading were effective for solving the problems. Moreover, an optimum value for the effective multiplication factor was also evaluated using reactivity coefficients. From the result, the maximum allowable value of the effective multiplication factor for a practical ADS can be set at 0.97. Consequently, a new core concept combining the buffer region and the two-zone loading was proposed base on the results of the parametric survey. (author)

  1. Study of the production of neutron-rich isotope beams issuing from fissions induced by fast neutrons

    International Nuclear Information System (INIS)

    Lau, Ch.

    2000-01-01

    This work is a contribution to the PARRNe project (production of radioactive neutron-rich isotopes). This project is based on the fission fragments coming from the fission of 238-uranium induced by fast neutrons. The fast neutron flux is produced by the collisions of deutons in a converter. Thick targets of uranium carbide and liquid uranium targets have been designed in order to allow a quick release of fission fragments. A device, able to trap on a cryogenic thimble rare gas released by the target, has allowed the production of radioactive nuclei whose half-life is about 1 second. This installation has been settled to different deuton accelerators in the framework of the European collaboration SPIRAL-2. A calibration experiment has proved the feasibility of fixing an ISOL-type isotope separator to a 15 MV tandem accelerator, this installation can provide 500 nA deutons beams whose energy is 26 MeV and be a valuable tool for studying fast-neutron induced fission. Zinc, krypton, rubidium, cadmium, iodine, xenon and cesium beams have been produced in this installation. The most intense beams reach 10000 nuclei by micro-coulomb for 26 MeV deutons. An extra gain of 2 magnitude orders can be obtained by using a more specific ion source and by increasing the thickness of the target. Another extra gain of 2 magnitude orders involves 100 MeV deutons

  2. Collimator optimization studies for the new MIT epithermal neutron beam

    International Nuclear Information System (INIS)

    Riley, K.J.; Ali, S.J.; Harling, O.K.

    2000-01-01

    A patient collimator has been designed for the epithermal neutron facility now being commissioned at MIT. Collimator performance both in and out of field was evaluated using the Monte Carlo code MCNP. A two piece design that can accommodate different circular field sizes will be manufactured using a composite lead, epoxy, boron and lithium mixture. (author)

  3. Neutron scattering study of yttrium iron garnet

    Science.gov (United States)

    Shamoto, Shin-ichi; Ito, Takashi U.; Onishi, Hiroaki; Yamauchi, Hiroki; Inamura, Yasuhiro; Matsuura, Masato; Akatsu, Mitsuhiro; Kodama, Katsuaki; Nakao, Akiko; Moyoshi, Taketo; Munakata, Koji; Ohhara, Takashi; Nakamura, Mitsutaka; Ohira-Kawamura, Seiko; Nemoto, Yuichi; Shibata, Kaoru

    2018-02-01

    The nuclear and magnetic structure and full magnon dispersions of yttrium iron garnet Y3Fe5O12 have been studied using neutron scattering. The refined nuclear structure is distorted to a trigonal space group of R 3 ¯ . The highest-energy dispersion extends up to 86 meV. The observed dispersions are reproduced by a simple model with three nearest-neighbor-exchange integrals between 16 a (octahedral) and 24 d (tetrahedral) sites, Ja a, Ja d, and Jd d, which are estimated to be 0.00 ±0.05 , -2.90 ±0.07 , and -0.35 ±0.08 meV, respectively. The lowest-energy dispersion below 14 meV exhibits a quadratic dispersion as expected from ferromagnetic magnons. The imaginary part of q -integrated dynamical spin susceptibility χ″(E ) exhibits a square-root energy dependence at low energies. The magnon density of state is estimated from χ″(E ) obtained on an absolute scale. The value is consistent with the single chirality mode for the magnon branch expected theoretically.

  4. Soft error rate simulation and initial design considerations of neutron intercepting silicon chip (NISC)

    Science.gov (United States)

    Celik, Cihangir

    -scale technologies. Prevention of SEEs has been studied and applied in the semiconductor industry by including radiation protection precautions in the system architecture or by using corrective algorithms in the system operation. Decreasing 10B content (20%of natural boron) in the natural boron of Borophosphosilicate glass (BPSG) layers that are conventionally used in the fabrication of semiconductor devices was one of the major radiation protection approaches for the system architecture. Neutron interaction in the BPSG layer was the origin of the SEEs because of the 10B (n,alpha) 7Li reaction products. Both of the particles produced have the capability of ionization in the silicon substrate region, whose thickness is comparable to the ranges of these particles. Using the soft error phenomenon in exactly the opposite manner of the semiconductor industry can provide a new neutron detection system based on the SERs in the semiconductor memories. By investigating the soft error mechanisms in the available semiconductor memories and enhancing the soft error occurrences in these devices, one can convert all memory using intelligent systems into portable, power efficient, directiondependent neutron detectors. The Neutron Intercepting Silicon Chip (NISC) project aims to achieve this goal by introducing 10B-enriched BPSG layers to the semiconductor memory architectures. This research addresses the development of a simulation tool, the NISC Soft Error Analysis Tool (NISCSAT), for soft error modeling and analysis in the semiconductor memories to provide basic design considerations for the NISC. NISCSAT performs particle transport and calculates the soft error probabilities, or SER, depending on energy depositions of the particles in a given memory node model of the NISC. Soft error measurements were performed with commercially available, off-the-shelf semiconductor memories and microprocessors to observe soft error variations with the neutron flux and memory supply voltage. Measurement

  5. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  6. Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

    Science.gov (United States)

    de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

    The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

  7. Monte Carlo neutronics analysis of the ANS reactor three-element core design

    International Nuclear Information System (INIS)

    Wemple, C.A.

    1995-01-01

    The advanced neutron source (ANS) is a world-class research reactor and experimental center for neutron research, currently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(fission) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. It was designed to be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. This paper summarizes the neutronics efforts at the Idaho National Engineering Laboratory in support of the development and analysis of the three-element core for the advanced conceptual design phase

  8. Design, construction and testing of a self-powered neutron detector

    International Nuclear Information System (INIS)

    Correa, R.F.

    1987-01-01

    The design, construction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties of detector and cable, gamma spectra induced in the detector through reactor irradiation, detector response as a function of neutron flux, current stability and reproductibility with the neutron flux. The gamma and neutron sensitivities were also evaluated, by means of thermoluminescent dosimeters and gold foils as references. The test results are presented and show that the detector response is reliable. The gamma and neutron sensitivities are in agreement with those found in the available literature. Nevertheless, a ceramic insulated cable should be employed for permanent use in a reactor. The tests were performed in a 100 KW TRIGA Mark I reactor at the Centro de Desenvolvimento da Tecnologia Nuclear of NUCLEBRAS, in Belo Horizonte, Brazil. (author) [pt

  9. Design constrution and testing of a self-powered neutron detector

    International Nuclear Information System (INIS)

    Correa, R.F.

    1987-01-01

    The design, contruction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties od detector and cable, gamma spectra induced in the detector through reactor irradiation, detector response as a function of neutron flux, current stability and reproductibility with the neutron flux. The gamma and neutron sensitivities were also evaluated, by means of thermoluminescent dosimeters and gold foils as references. The test results are presented and show that the detector response is reliable. The gamma and neutron sensitivities are in agreement with those found in the available literature. Neverthe less, a ceramic insulated cable should be employed for permanent use in a reactor. The tests were perfomance in a 100 kW TRIGA Mark I reactor at the Centro de Desenvolvimento da Tecnologia Nuclear of NUCLEBRAS,in Belo Horizonte, Brazil. (Author) [pt

  10. How should the JAERI neutron source be designed?

    International Nuclear Information System (INIS)

    Watanabe, Noboru

    1996-01-01

    The importance of a next-generation neutron source in JAERI is discussed. The feasibility and the performances of three types of neutron sources, namely continuous wave spallation source (CWSS), long-pulse spallation source (LPSS) and short-pulse spallation source (SPSS), are compared based on a proposed JAERI accelerator, a superconducting (SC) proton linac (1-1.5 GeV, 25-16 mA in peak current, finally CW). How to realize one of the world's best neutron source using such a linac with a modest beam-current and what type of neutron source is the best for such a linac are the most important current problems. Since the accelerator is not favorable for LPSS due to a lower peak current and there exist serious technical problems for a CWSS target, a short-pulse spallation source would be the best candidate to realize a 5 MW-class SPSS like ESS, provided that the H - -injection to a compressor ring over a long pulse duration (>2 ms) is feasible. (author)

  11. Methodology for worker neutron exposure evaluation in the PDCF facility design

    International Nuclear Information System (INIS)

    Scherpelz, R. I.; Traub, R. J.; Pryor, K. H.

    2004-01-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y -1 for the whole body and 100 mSv y -1 for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were

  12. Thermo-mechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine

    International Nuclear Information System (INIS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2009-01-01

    Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 deg. C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  13. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  14. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  15. Low temperature and neutron physics studies

    International Nuclear Information System (INIS)

    Shull, C.G.

    1989-01-01

    A search for a novel coupling interaction between the Pendelloesung periodicity which is formed in a diffracting crystal and the Larmor precession of neutrons in a magnetic field has been carried out. This interaction is expected to exhibit a resonant behavior when the two spatial periodicities become matched upon scanning the magnetic field being applied to the crystal. Observations on a diffracting, perfect crystal of silicon with neutrons of wavelength 1 Angstrom show the expected resonant action but some discrepancy between the observed magnitude of the resonance effects remains for interpretation. 16 refs

  16. Neutronic analysis of graphite-moderated solid breeder design for INTOR

    International Nuclear Information System (INIS)

    Jung, J.; Abdou, M.A.

    1981-01-01

    An in-depth analysis of the INTOR tritium-production-blanket design is presented. A ternary system of solid silicate breeder, lead neutron multiplier, and graphite moderator is explored primary from safety and blanket tritium-inventory considerations. Lithium-silicate (Li 2 SiO 3 ) breeder systems are studied along with water (H 2 O/D 2 O) and Type 316 stainless steel as coolant and structural material, respectively. The analysis examines the neutronics effects on tritium-production regarding: (1) coolant choice; (2) moderator choice; (3) moderator location; (4) multiplier thickness; (5) 6 Li enrichment; and (6) 6 Li burnup. The tritium-breeding-blanket modules are located at the top, outboard, and bottom (outer) parts of the torus, resulting in a breeding coverage of approx. 60% at the first-wall surface. It is found that the reference INTOR design yields, based on a three-dimensional analysis, a net tritium breeding ratio (BR) of approx. 0.65 at the beginning of reactor operation, satisfying the design criterion of BR > 0.6

  17. Analysis of a shield design for a DT neutron generator test facility.

    Science.gov (United States)

    Chichester, D L; Pierce, G D

    2007-10-01

    Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

  18. 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

  19. Magnetic particles studied with neutron depolarization and small-angle neutron scattering

    International Nuclear Information System (INIS)

    Rosman, R.

    1991-01-01

    Materials containing magnetic single-domain particles, referred to as 'particulate media', have been studied using neutron depolarization (ND) and small-angle neutron scattering (SANS). In a ND experiment the polarization vector of a polarized neutron beam is analyzed after transmission through a magnetic medium. Such an analysis in general yields the correlation length of variations in magnetic induction along the neutron path (denoted 'magnetic correlation length'), mean orientation of these variations and mean magnetic induction. In a SANS experiment, information about nuclear and magnetic inhomogeneities in the medium is derived from the broadening of a generally unpolarized neutron beam due to scattering by these inhomogeneities. Spatial and magnetic microstructure of a variety of particulate media have been studied using ND and/or SANS, by determination of the magnetic or nuclear correlation length in these media in various magnetic states. This thesis deals with the ND theory and its application to particulate media. ND and SANS experiments on a variety of particulate media are discussed. (author). 178 refs., 97 figs., 8 tabs

  20. Analysis of phase velocity designing on superconducting section of proton Linac for spallation neutron source

    International Nuclear Information System (INIS)

    Ouyang Huafu; Xu Taoguang; Yu Qingchang; Guan Xialing; Luo Zihua

    2001-01-01

    A preliminary design of superconducting section of proton linac for spallation neutron source is made, which includes the design and optimization of the cavity shape and the architecture design of the superconducting section. In addition, the choice of the cell number of the superconducting cavity, the value of the geometric β G , the optimization principles of cavity and the beam dynamic properties are discussed

  1. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    Setiadipura, T.; Zuhair; Irwanto, D.

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  2. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  3. Design of a system for neutrons dosimetry; Diseno de un sistema para dosimetria de neutrones

    Energy Technology Data Exchange (ETDEWEB)

    Ceron, P.; Rivera, T. [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Legaria No. 694, Col. Irrigacion, 11500 Mexico D. F. (Mexico); Paredes G, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Sanchez, A. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Vega C, H. R., E-mail: victceronr@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    At the present time diverse systems of detection of neutrons exist, as proportional counters based on BF{sub 3}, He{sub 3} and spectrometers of Bonner spheres. However, the cost and the complexity of the implementation of these systems put them far from the reach for dosimetric purposes. For these reasons a system of neutrons detection composed by a medium paraffin moderator that forms a 4π (spheres) arrangement and of several couples of thermoluminescent dosimeters TLD 600/TLD 700. The response of the system presents a minor repeatability to 5% in several assays when being irradiated with a {sup 239}PuBe source and a deviation of 13.8% in the Tl readings of four different spheres. The calibration factor of the system with regard to the neutrons source which was of 56.2 p Sv/nc also was calculated. These detectors will be used as passive monitors of photoneutrons in a radiotherapy room with lineal accelerator of high energy. (Author)

  4. Design and experimental tests of a novel neutron spin analyzer for wide angle spin echo spectrometers

    Energy Technology Data Exchange (ETDEWEB)

    Fouquet, Peter; Farago, Bela; Andersen, Ken H.; Bentley, Phillip M.; Pastrello, Gilles; Sutton, Iain; Thaveron, Eric; Thomas, Frederic [Institut Laue-Langevin, BP 156, F-38042 Grenoble Cedex 9 (France); Moskvin, Evgeny [Helmholtzzentrum Berlin, Glienicker Strasse 100, D-14109 Berlin (Germany); Pappas, Catherine [Helmholtzzentrum Berlin, Glienicker Strasse 100, D-14109 Berlin (Germany); Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2009-09-15

    This paper describes the design and experimental tests of a novel neutron spin analyzer optimized for wide angle spin echo spectrometers. The new design is based on nonremanent magnetic supermirrors, which are magnetized by vertical magnetic fields created by NdFeB high field permanent magnets. The solution presented here gives stable performance at moderate costs in contrast to designs invoking remanent supermirrors. In the experimental part of this paper we demonstrate that the new design performs well in terms of polarization, transmission, and that high quality neutron spin echo spectra can be measured.

  5. Development of new methods for studying nanostructures using neutron scattering

    International Nuclear Information System (INIS)

    Pynn, Roger

    2016-01-01

    The goal of this project was to develop improved instrumentation for studying the microscopic structures of materials using neutron scattering. Neutron scattering has a number of advantages for studying material structure but suffers from the well-known disadvantage that neutrons' ability to resolve structural details is usually limited by the strength of available neutron sources. We aimed to overcome this disadvantage using a new experimental technique, called Spin Echo Scattering Angle Encoding (SESAME) that makes use of the neutron's magnetism. Our goal was to show that this innovation will allow the country to make better use of the significant investment it has recently made in a new neutron source at Oak Ridge National Laboratory (ORNL) and will lead to increases in scientific knowledge that contribute to the Nation's technological infrastructure and ability to develop advanced materials and technologies. We were successful in demonstrating the technical effectiveness of the new method and established a baseline of knowledge that has allowed ORNL to start a project to implement the method on one of its neutron beam lines.

  6. VESUVIO: a novel instrument for performing spectroscopic studies in condensed matter with eV neutrons at the ISIS facility

    Science.gov (United States)

    Senesi, R.; Andreani, C.; Bowden, Z.; Colognesi, D.; Degiorgi, E.; Fielding, A. L.; Mayers, J.; Nardone, M.; Norris, J.; Praitano, M.; Rhodes, N. J.; Stirling, W. G.; Tomkinson, J.; Uden, C.

    2000-03-01

    The VESUVIO project aims to provide unique prototype instrumentation at the ISIS-pulsed neutron source and to establish a routine experimental and theoretical program in neutron scattering spectroscopy at eV energies. This instrumentation will be specifically designed for high momentum, (20 Å-11 eV) inelastic neutron scattering studies of microscopic dynamical processes in materials and will represent a unique facility for EU researchers. It will allow to derive single-particle kinetic energies and single-particle momentum distributions, n(p), providing additional and/or complementary information to other neutron inelastic spectroscopic techniques.

  7. Neutron-proton bremsstrahlung studies using the white neutron source at the LAMPF/WNR

    International Nuclear Information System (INIS)

    Wender, S.A.; Nelson, R.O.; Schillaci, M.E.; Blann, M.

    1990-01-01

    Nucleon-nucleon bremsstrahlung is a few-body radiative process that provides insight into several areas of nuclear physics. It is one of the simplest systems for studying the off-shell behavior of the nucleon-nucleon potential. The physics involved in neutron-proton bremsstrahlung (NPB) is significantly different from that of proton-proton bremsstrahlung (PPB). In particular, NPB cross sections are much larger than PPB cross sections because NPB allows E1 radiation, and the contribution to the cross section from the meson exchange currents has been calculated to be as large as the contributions from external radiation. To date there have been essentially four NPB experiments. These measurements have covered only a small part of the available phase space. A major experimental problem in performing these measurements has been the lack of a suitable intense, high-energy neutron beam. We are planning a measurement of the NPB cross section using the white neutron source at the WNR target area at the LAMPF accelerator. We plant to implement the experiment in three phases. In this first state, we shall measure inclusive hard-photon production using a multi-element gamma-ray telescope that is insensitive to neutrons. In the second phase, we shall measure the bremsstrahlung gamma-rays in coincidence with recoil protons. In the last phase, we shall detect the scattered neutrons in coincidence with the recoil protons and gamma rays. 8 refs., 6 figs

  8. Neutron Crystallography for the Study of Hydrogen Bonds in Macromolecules

    Directory of Open Access Journals (Sweden)

    Esko Oksanen

    2017-04-01

    Full Text Available Abstract: The hydrogen bond (H bond is one of the most important interactions that form the foundation of secondary and tertiary protein structure. Beyond holding protein structures together, H bonds are also intimately involved in solvent coordination, ligand binding, and enzyme catalysis. The H bond by definition involves the light atom, H, and it is very difficult to study directly, especially with X-ray crystallographic techniques, due to the poor scattering power of H atoms. Neutron protein crystallography provides a powerful, complementary tool that can give unambiguous information to structural biologists on solvent organization and coordination, the electrostatics of ligand binding, the protonation states of amino acid side chains and catalytic water species. The method is complementary to X-ray crystallography and the dynamic data obtainable with NMR spectroscopy. Also, as it gives explicit H atom positions, it can be very valuable to computational chemistry where exact knowledge of protonation and solvent orientation can make a large difference in modeling. This article gives general information about neutron crystallography and shows specific examples of how the method has contributed to structural biology, structure-based drug design; and the understanding of fundamental questions of reaction mechanisms.

  9. Design features and operating experiences of neutron measurement system for HANARO

    International Nuclear Information System (INIS)

    Kim, Young Ki; Choi, Young San

    1999-02-01

    This paper discusses unique mechanical and electrical design features of neutron measurement system for HANARO and its operating experiences. Some unexpected problems and misbehaviors during installation and commissioning are briefly introduced. Engineering approaches and procedures in order to solve the problems we are described in priority. It has been proved, through years of operation that the wide range neutron measurement system for HANARO has a good capability of providing the stable and reliable neutron flux signal for reactor control and reactor protection. I hopefully expect that the engineering solutions suggested in this report could be a good reference to the other applications. (Author). 12 refs., 6 tabs., 23 figs

  10. Fundamental design of systems and facilities for cold neutron source in the Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Soo; Jeong, H. S.; Kim, Y. K.; Wu, S. I

    2006-01-15

    The CNS(Cold Neutron Source) development project has been carried out as the partial project of the reactor utilization R and D government enterprise since 2003. In the advantage of lower energy and long wave length for the cold neutron, it can be used with the essential tool in order to investigate the structure of protein, amino-acid, DNA, super lightweight composite and advanced materials in the filed of high technology. This report is mainly focused on the basic design of the systems and facilities for the HANARO cold neutron source, performed during the second fiscal project year.

  11. Fundamental design of systems and facilities for cold neutron source in the Hanaro

    International Nuclear Information System (INIS)

    Kim, Bong Soo; Jeong, H. S.; Kim, Y. K.; Wu, S. I.

    2006-01-01

    The CNS(Cold Neutron Source) development project has been carried out as the partial project of the reactor utilization R and D government enterprise since 2003. In the advantage of lower energy and long wave length for the cold neutron, it can be used with the essential tool in order to investigate the structure of protein, amino-acid, DNA, super lightweight composite and advanced materials in the filed of high technology. This report is mainly focused on the basic design of the systems and facilities for the HANARO cold neutron source, performed during the second fiscal project year

  12. Plant protection system optimization studies to mitigate consequences of large breaks in the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Khayat, M.I.; March-Leuba, J.

    1993-01-01

    This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the advanced neutron source (ANS) reactor. The ANS is a new basic and applied research facility based on a powerful steady-state research reactor that provides beams of neutrons for measurements and experiments in the field of material science and engineering, biology, chemistry, material analysis, and nuclear science. To achieve the high neutron fluxes for these state-of-the-art experiments, the ANS design has a very high power density core (330 MW fission with an active volume of 67.6 ell) surrounded by a large heavy-water reflector, where most neutrons are moderated. This design maximizes the number of neutrons available for experiments but results in a low heat capacity core that creates unique challenges to the design of the plant protection system

  13. Trojan Horse Method for neutrons-induced reaction studies

    Science.gov (United States)

    Gulino, M.; Asfin Collaboration

    2017-09-01

    Neutron-induced reactions play an important role in nuclear astrophysics in several scenario, such as primordial Big Bang Nucleosynthesis, Inhomogeneous Big Bang Nucleosynthesis, heavy-element production during the weak component of the s-process, explosive stellar nucleosynthesis. To overcome the experimental problems arising from the production of a neutron beam, the possibility to use the Trojan Horse Method to study neutron-induced reactions has been investigated. The application is of particular interest for reactions involving radioactive nuclei having short lifetime.

  14. Study on the transmission efficiency of curved neutron guide

    International Nuclear Information System (INIS)

    Wang Hongli; Zhang Li; Guo Liping; Yang Tonghua; Zhao Zhixiang

    2004-01-01

    Monte-Carlo simulation program NGT2002 is used to study the transmission efficiency of curved neutron guide from character wavelength, film reflectivity, film material, geometry adjustment error, gap between guides and guide fabricate error, the authors get the transmission efficiency curves of the Ni, supper mirror curved neutron guides, also we have a discuss of how to choose the curved neutron guide's character wavelength. By the simulation results, the authors determine the proper film reflectivity value, guide horizontal geometry adjustment error range, optimized gap value between guide elements and guide width fabricate geometry error range. (authors)

  15. Neutron scattering studies of the heavy Fermion superconductors

    International Nuclear Information System (INIS)

    Goldman, A.I.

    1985-01-01

    Recent neutron scattering measurements of the heavy Fermion superconductors are described. Those materials offer an exciting opportunity for neutron scattering since the f-electrons, which couple directly to magnetic scattering measurements, seem to be the same electrons which form the superconducting state below T/sub c/. In addition, studies of the magnetic fluctuations in these, and other heavy Fermion systems, by inelastic magnetic neutron scattering can provide information about the nature of the low temperature Fermi liquid character of these novel compounds

  16. Backtracing neutron analysis in the fusion-fission dynamics study

    International Nuclear Information System (INIS)

    Brennand, E. de Goes; Hanappe, F.; Stuttge, L.

    2001-01-01

    A new method for the analysis of multi parametric experimental data is used in the study of the dynamics of the fission process for the compound system 126 Ba. We apply this method to obtain the correlation between thermal energy related to the neutron total multiplicity and the correlation between pre-scission neutron and pos-scission neutron multiplicities. The results obtained are interpreted into the framework of a dynamical model. From this interpretation we have access to the following information: the friction intensity which drives the dynamical evolution of the system; the initial deformation of the compound system; the barrier evolution with temperature and angular momentum, and fission times. (author)

  17. Experimental study of modification of neutron spectrum using filters

    International Nuclear Information System (INIS)

    Kobayashi, H.; Matsubayashi, M.; Brenizer, J.S. Jr.; Lindsay, J.T.

    1996-01-01

    Filter effects for continuum thermal neutron beams were experimentally studied by means of an effective energy. Be, Bi and Pb were used as filter materials to examine the energy shift of the spectrum. It was found that the effective energy of a thermal neutron beam is easily lowered into a sub-thermal region (down to 10 meV) by use of a filter system without any cooling system. The effectiveness and its applicability will be discussed in this study. (orig.)

  18. The dynamics of physisorbed layers studied by neutron scattering

    International Nuclear Information System (INIS)

    Nielsen, M.; McTague, J.P.

    1978-01-01

    We discuss the neutron scattering technique applied to the study of adsorbed thin films. Despite the fact that neutrons are scattered very weakly by surfaces, recent studies have shown that both structural and dynamical information can be obtained even for submonolayer coverages. Results will be shown for films of Ar, D 2 , H 2 , and O 2 adsorbed on (001) surfaces of graphite and for H 2 molecules adsorbed on activated alumina. (orig.) [de

  19. Design basis neutronics calculations for NRU-LOCA experiments

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Jenquin, U.P.; McNair, G.W.; Perry, R.T.; Trapp, T.J.; Zimmerman, M.G.

    1979-08-01

    The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described

  20. Shielding calculations for the design of neutron radiography facility around PARR

    International Nuclear Information System (INIS)

    Ashraf, M.M.; Khan, A.R.

    1989-06-01

    Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)

  1. Design and fabrication of a self-powered neutron detector

    International Nuclear Information System (INIS)

    Garcia Garcia, Florencio.

    1979-01-01

    Self powered neutron detectors are becoming more and more popular in reactor instrumentation. A fast response detector of this type was made at the Reactor Division, La Reina Nuclear Center in Santiago. Cobalt wire was the emitter, teflon the insulator and a stainless steel tubing was the collector. The overall dimensions of the detector are 6 mms diameter and 700 mms length. The irradiation tests, carried out at the Center's 5 Mw research reactor showed a very reasonably linear relation between current supplied by the detector and thermal neutron flux, over a range extending from 10 10 to 10 13 n/cm 2 x seg. These tests also showed a good agreement between calculated and measured current. The models used for the calculation of current are fully explained and they include some improvements over those that have been published recently. An important conclusion for the case of the cobalt detectors is that the wire's diameter must be at least 1 mm. in order to have a neutron induced current bigger than the parasitic components generated by indirect processes. Calculations for other emitters such as vanadium, silver and rhodum are also included. (EC)

  2. neutron detector for in-beam studies

    International Nuclear Information System (INIS)

    Schmitt, R.P.; Nebbia, G.; Fabris, D.; Natowitz, J.B.; Utsunomiya, H.; Wada, R.

    1987-01-01

    Flexible, high-geometry detection systems are indispensable in unraveling the complexities of the contributing reaction mechanisms in medium energy heavy-ion collisions. In preparation for the K500 cyclotron, which will come on-line in 1987, they are constructing a 4π neutron ball. Like the fission neutron tanks first constructed more than three decades ago, the neutron ball consists of a large volume (approximately 1700 1) of Gd-doped liquid scintillator. However, the ball is distinguished from these systems in its relatively large scattering chamber and modular design. The design features and the expected performance of the ball will be described. They will also report on the current status of the project

  3. Design and fabrication of a novel self-powered solid-state neutron detector

    Science.gov (United States)

    LiCausi, Nicholas

    key advantages of the novel devices are theoretical neutron detection efficiency of ˜48%, a self-passivating structure that reduces leakage current and detector operation with no bias resulting in extremely low device noise. Processes required to fabricate the 3D type detector were explored and developed in this thesis. The detector capacitance and processing steps have been simulated with MEDICI and TSuprem-4, respectively. Lithography masks were then designed using Cadence. The fabrication process development was conducted in line with standard CMOS grade integrated circuit processing to allow for simple integration with existing fabrication facilities. A number of new processes were developed including the low pressure chemical vapor deposition of conformal boron films using diborane on very high aspect-ratio trenches and holes. Development also included methods for "wet" chemical etching and "dry" reactive ion etching of the deposited boron films. Fabricated detectors were characterized with the transmission line method, 4-point probe, I-V measurements and C-V measurements. Finally the detector response to thermal neutrons was studied. Characterization has shown significant reduction in reverse leakage current density to ˜8x10-8 A/cm2 (nearly 4 orders of magnitude over the previously published data). Results show that the fabrication process developed is capable of producing efficient (˜22.5%) solid-state thermal neutron detectors.

  4. The AECL study for an intense neutron - generator (technical details)

    Energy Technology Data Exchange (ETDEWEB)

    Bartholomew, G A; Tunnicliffe, P R

    1966-07-01

    The AECL study for an intense neutron-generator has been in progress for two years. Recently the scientific and technical details and the conceptual designs were compiled in a report supporting proposals addressed to AECL's Board of Directors for further work. The compilation is being issued in this form to permit further discussion of the technical aspects. However readers are asked to appreciate that it was written primarily for an AECL audience, and specifically that those chapters giving tentative information about costs, the rate of investment and similar items have been omitted or modified, many references have been made to interim internal reports in order to complete the local documentation, but these references do not imply that the reports themselves can be made generally available. (author)

  5. The AECL study for an intense neutron - generator (technical details)

    Energy Technology Data Exchange (ETDEWEB)

    Bartholomew, G.A.; Tunnicliffe, P.R

    1966-07-01

    The AECL study for an intense neutron-generator has been in progress for two years. Recently the scientific and technical details and the conceptual designs were compiled in a report supporting proposals addressed to AECL's Board of Directors for further work. The compilation is being issued in this form to permit further discussion of the technical aspects. However readers are asked to appreciate that it was written primarily for an AECL audience, and specifically that those chapters giving tentative information about costs, the rate of investment and similar items have been omitted or modified, many references have been made to interim internal reports in order to complete the local documentation, but these references do not imply that the reports themselves can be made generally available. (author)

  6. The AECL study for an intense neutron - generator (technical details)

    International Nuclear Information System (INIS)

    Bartholomew, G.A.; Tunnicliffe, P.R.

    1966-01-01

    The AECL study for an intense neutron-generator has been in progress for two years. Recently the scientific and technical details and the conceptual designs were compiled in a report supporting proposals addressed to AECL's Board of Directors for further work. The compilation is being issued in this form to permit further discussion of the technical aspects. However readers are asked to appreciate that it was written primarily for an AECL audience, and specifically that those chapters giving tentative information about costs, the rate of investment and similar items have been omitted or modified, many references have been made to interim internal reports in order to complete the local documentation, but these references do not imply that the reports themselves can be made generally available. (author)

  7. Conceptual Design of a 14-MeV D-T Neutron Source for Material Inspection

    International Nuclear Information System (INIS)

    Kim, Jin-Choon; Oh, Byung-Hoon

    2007-01-01

    There is a worldwide need for the efficient inspection of cargo containers at airports, seaports and border crossings. And there is also a growing need for nondestructive inspection of metal objects such as airplane parts. The limitations of X-ray systems for the detection of explosives, drugs, and thick metal structures have stimulated interest in neutron radiograph or tomography. The weak link in such applications is the neutron source. The ideal neutron source should provide a high intensity, high-energy for sufficient penetration and activation, a reliable long-term operation, and a monoenergetic neutron beam. In this paper, we describe a conceptual design of a DT fusion neutron source (monoenergetic 14 MeV neutron generator) which satisfies the fore-mentioned requirements. The current design is based upon the actually proven system using the drive-in target principle. The design is versatile enough to accommodate various applications, ranging from material inspection and explosive interrogation to medical probing and cancer treatment

  8. MCNPX Simulation Study of STRAW Neutron Detectors - Summary Paper

    International Nuclear Information System (INIS)

    Mukhopadhyay, Sanjoy; Maurer, Richard; Mitchell, Stephen

    2010-01-01

    A novel prototype fission meter is being designed at National Security Technologies, LLC, using a thin uniform coating (only 1 micron thick) of 10 B as a neutron converter inside a large array of thin (4 mm diameter) copper tubes. The copper tubes are only 2 mils thick, and each holds the stretched anode wire under tension and high voltage. The tubes are filled with proportional counter gas (a mixture of 90%/10% of Ar/CO 2 ). The tubes operate in proportional counter mode and attract mobile charged particles (α's) created in the nuclear interaction 10 B(n, 4 He) 7 Li. However, a single tube has about 1/7th the sensitivity of a 3 He tube. Modeling is required to determine if enough such tubes could be placed in a neutron detection assembly of the current size to give comparable sensitivity to 3 He. Detectors lined with 10 B lie between 3 He and 10 BF 3 proportional counters and fission chambers in terms of neutron detection efficiency and gamma ray insensitivity. The mean free path of thermal neutrons in 10 B is about 18 (micro)m. It takes about 60 (micro)m of 10 B layer to completely stop thermal neutrons, but the energetic α-particles generated in the reaction have a range of only 3.3 (micro)m in 10 B environment - hence the thin layer of boron coating on the copper tube. The prototype design is shown in Figure 1. It consists of two panels of three staggered rows of 500-mm-long, 4-mm-diameter straws, with 20 in each row, embedded in 30-mm-thick high density polyethylene (HDPE). The project demonstrates a new application of thin neutron and gamma converter technique (1 micron thin 10 B coated copper tube). It exploits fast timing from multiple straw detectors to count multiplicity of both gamma and neutrons from fissioning materials. The objective is to find a near-term replacement of 3 He gas in neutron detection and measurement (with a very large neutron detection area). All the solid-state detectors developed thus far are small and inefficient. The thin size

  9. Study on neutron beam probe. Study on the focused neutron beam

    Energy Technology Data Exchange (ETDEWEB)

    Kotajima, Kyuya; Suzuki, K.; Fujisawa, M.; Takahashi, T.; Sakamoto, I. [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Wakabayashi, T.

    1998-03-01

    A monoenergetic focused neutron beam has been produced by utilizing the endoenergetic heavy ion reactions on hydrogen. To realize this, the projectile heavy ion energy should be taken slightly above the threshold energy, so that the excess energy converted to the neutron energy should be very small. In order to improve the capability of the focused neutron beam, some hydrogen stored metal targets have also been tested. Separating the secondary heavy ions (associated particles) from the primary ions (accelerated particles) by using a dipole magnet, a rf separator, and a particle identification system, we could directly count the produced neutrons. This will leads us to the possibility of realizing the standard neutron field which had been the empty dream of many neutron-related researchers in the world. (author)

  10. Experimental design technique applied to the validation of an instrumental Neutron Activation Analysis procedure

    International Nuclear Information System (INIS)

    Santos, Uanda Paula de M. dos; Moreira, Edson Gonçalves

    2017-01-01

    In this study optimization of procedures and standardization of Instrumental Neutron Activation Analysis (INAA) method were carried out for the determination of the elements bromine, chlorine, magnesium, manganese, potassium, sodium and vanadium in biological matrix materials using short irradiations at a pneumatic system. 2 k experimental designs were applied for evaluation of the individual contribution of selected variables of the analytical procedure in the final mass fraction result. The chosen experimental designs were the 2 3 and the 2 4 , depending on the radionuclide half life. Different certified reference materials and multi-element comparators were analyzed considering the following variables: sample decay time, irradiation time, counting time and sample distance to detector. Comparator concentration, sample mass and irradiation time were maintained constant in this procedure. By means of the statistical analysis and theoretical and experimental considerations, it was determined the optimized experimental conditions for the analytical methods that will be adopted for the validation procedure of INAA methods in the Neutron Activation Analysis Laboratory (LAN) of the Research Reactor Center (CRPq) at the Nuclear and Energy Research Institute (IPEN /CNEN-SP). Optimized conditions were estimated based on the results of z-score tests, main effect, interaction effects and better irradiation conditions. (author)

  11. Experimental design technique applied to the validation of an instrumental Neutron Activation Analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Uanda Paula de M. dos; Moreira, Edson Gonçalves, E-mail: uandapaula@gmail.com, E-mail: emoreira@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    In this study optimization of procedures and standardization of Instrumental Neutron Activation Analysis (INAA) method were carried out for the determination of the elements bromine, chlorine, magnesium, manganese, potassium, sodium and vanadium in biological matrix materials using short irradiations at a pneumatic system. 2{sup k} experimental designs were applied for evaluation of the individual contribution of selected variables of the analytical procedure in the final mass fraction result. The chosen experimental designs were the 2{sup 3} and the 2{sup 4}, depending on the radionuclide half life. Different certified reference materials and multi-element comparators were analyzed considering the following variables: sample decay time, irradiation time, counting time and sample distance to detector. Comparator concentration, sample mass and irradiation time were maintained constant in this procedure. By means of the statistical analysis and theoretical and experimental considerations, it was determined the optimized experimental conditions for the analytical methods that will be adopted for the validation procedure of INAA methods in the Neutron Activation Analysis Laboratory (LAN) of the Research Reactor Center (CRPq) at the Nuclear and Energy Research Institute (IPEN /CNEN-SP). Optimized conditions were estimated based on the results of z-score tests, main effect, interaction effects and better irradiation conditions. (author)

  12. Nuclear structure studies via neutron interactions

    International Nuclear Information System (INIS)

    Carlton, R.F.

    1990-03-01

    Research performed consisted of: refinement of previous analysis of high resolution total cross sections for n + 40 Ar in an effort to remove some ambiguities in J π assignments and completion of two papers dealing with this analysis and a comparison theoretical treatment of the associated scattering functions and R-functions; extension of the analysis of neutron total cross section data on 48 Ca to 3.5 MeV in neutron energy and modeling of the results with a dispersive optical model based on parameters from 40 Ca scattering data; attempted improvement of spin and parity assignments for data on 122 Sn and determination of external R-function parameters; development of a graphical interface, coupled with a code for calculation of R-matrix based total cross sections and parameter minimization, for an MS-DOS-based microcomputer. 11 refs., 13 figs

  13. Neutron diffraction studies of magnetic materials

    International Nuclear Information System (INIS)

    James, W.J.

    1987-01-01

    The ability of neutron diffraction in determining the nature and extent of magnetic ordering is illustrated for the intermetallic compounds, Y/sub 6/(Fe,Mn)/sub 23/ and ErFe/sub 3/. Substitution with other 3d transition metals influences the Fe-Fe exchange forces such as to alter, sometimes considerably, the magnetic properties, e.g., local site magnetic anisotropies in Er)Fe,Ni)/sub 3/ and thermal expansion anomalies in the R/sub 2/)Fe,Co)/sub 14/B compounds. When the 3d atoms are near neighbors in the periodic chart, their nuclear scattering lengths for neutrons are sufficiently different to permit the detection of preferential occupation of the several nonequivalent crystallographic 3d metal sites, i.e., atomic ordering, present in the R/sub 6/M/sub 23/, and R/sub 2/Fe/sub 14/B structures

  14. Study of tunneling time with Larmor clock and neutron absorption

    International Nuclear Information System (INIS)

    Hino, M.; Tasaki, S.; Ebisawa, T.; Kawai, T.; Utsuro, M.; Achiwa, N.

    2001-01-01

    Tunnel effect is one of the most typical quantum mechanical phenomena which cannot be understood in the classical physics. Though the tunnel phenomenon itself is precisely defined on the basis of quantum mechanics, tunneling time, the time for a particle to pass a tunnel barrier, has been a controversial issue because time is a parameter to show the rate of change of physical phenomena in quantum mechanics but has no corresponding quantum operator. In the present study, Larmor precession of the neutrons passing a Permalloy45 (PA) thin film was measured as a function of neutron incident angle to the film by using neutron spin-echo instrument at KUR and neutron interferometer at JRR-3. Results are compared with a calculation based on one-dimensional Schroedinger equation. The agreement between the experiment and the plane-wave simulation is very good which means that the neutrons are not described in particle picture here, and the Larmor time defined as number of spin precession divided by angular velocity no more represents the time for a particle to pass the barrier although it is a kind of the phase time. In order to emphasize the particle picture, effects of neutron absorption were considered theoretically. Larmor precession passing through a Fabry-Perot magnetic thin film which has two potential barriers (quantum well) for up-spin neutrons were measured for two cases that the film is hot neutron absorptive (PA-Ge-PA) and strongly absorptive (Se-Ge/Gd-Se). Here Ge, Gd and Se represent germanium, gadolinium and Supersendust, respectively. While down-spin neutrons feel only a small potential barrier. Spin-dependent reflectivity and transmission of the Fabry-Perot magnetic films were also measured as functions of the neutron incident angle to the film. Experimental results of the non-absorptive film show that the neutron spin precession cannot be treated as the classical motion of a magnetic moment feeling torque under applied magnetic field like the Larmor

  15. Thermal-hydraulic design concept of the solid-target system of spallation neutron source

    International Nuclear Information System (INIS)

    Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K.

    2001-01-01

    In relation to thermal-hydraulic design of the N-Arena solid-target system of the JHF project, heat transfer experiments were performed to obtain experimental data systematically on heat transfer coefficient and CHF for vertical upward and horizontal flows in a thin rectangular channel simulating a coolant channel of the proposed spallation neutron source. Thermal-hydraulic correlations which can be used for design calculations were proposed based on the obtained data. Finally tentative results of feasibility study on maximum beam power which could be attained with a solid target were presented. The result indicated that the condition for the onset of nucleate boiling is the most significant limiting factor to the maximum beam power. (author)

  16. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

    Pond, R.B.; Hanan, N.A.; Matos, J.E.

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  17. MCNP-REN: a Monte Carlo tool for neutron detector design

    International Nuclear Information System (INIS)

    Abhold, M.E.; Baker, M.C.

    2002-01-01

    The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions

  18. Neutronic scoping studies for the tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

    1976-02-01

    One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

  19. Spin dynamics above the Curie temperature studied by neutron scattering

    International Nuclear Information System (INIS)

    Steinsvoll, O.; Riste, T.

    1986-01-01

    Neutron scattering can in principle give information about magnetic fluctuations over the entire atomic space and time domain. The weakness of the neutron-matter interaction renders this information undistorted by the neutron probe, but at the same time puts intensity limitations on the method. A considerable number of studies on the magnetism of 3d metals have been performed at some of the larger reactor laboratories. In the regions of overlap the experimental results from the different laboratories are consistent, but the interpretations are along different lines. Among the controversial issues are itinerancy versus localization, the degree of order above T C . In our talk we shall give an introduction to the neutron scattering method, including some of the sophisticated polarized beam methods. In the rest of the talk we shall review recent experimental results and some of the theoretical models used in their interpretation. (orig.)

  20. Studies Of The (n, γ) Reaction With A Neutron Monochromator

    Energy Technology Data Exchange (ETDEWEB)

    Kane, W. R.; Gardner, D.; Brown, T.; Kevey, A.; Mateosian, E. der; Emery, G. T.; Gelletly, W.; Mariscotti, M. A.J. [Brookhaven National Laboratory, Upton, Long Island, NY (United States); Schröder, I. [National Bureau of Standards, Washington, DC (United States)

    1969-11-15

    A crystal diffraction neutron monochromator has been constructed specifically for studies of the(n, γ) reaction. This equipment plays a complementary role to that of time-of-flight devices in providing a neutron beam with a full duty cycle at a given energy. This feature and the small target size, large geometrical efficiency for y-ray detection, and negligible fast neutron background afford advantages for certain classes of experiments. The useful energy range extends from 0.01 to 20 eV. Novel features of the equipment include a complete reliance upon precision angle encoders for setting arm and crystal angles, the employment of a liquid shield to facilitate the extraction of the diffracted neutron beam, and the use of air bearings to provide for the motion of the target, detection devices, and associated shielding. Results obtained on low energy resonances of {sup 139}La, {sup 189}Os, and {sup 235}U will be presented. (author)

  1. BEAM-LOSS DRIVEN DESIGN OPTIMIZATION FOR THE SPALLATION NEUTRON SOURCE (SNS) RING.

    Energy Technology Data Exchange (ETDEWEB)

    WEI,J.; BEEBE-WANG,J.; BLASKIEWICZ,M.; CAMERON,P.; DANBY,G.; GARDNER,C.J.; JACKSON,J.; LEE,Y.Y.; LUDEWIG,H.; MALITSKY,N.; RAPARIA,D.; TSOUPAS,N.; WENG,W.T.; ZHANG,S.Y.

    1999-03-29

    This paper summarizes three-stage design optimization for the Spallation Neutron Source (SNS) ring: linear machine design (lattice, aperture, injection, magnet field errors and misalignment), beam core manipulation (painting, space charge, instabilities, RF requirements), and beam halo consideration (collimation, envelope variation, e-p issues etc.).

  2. Beam-Loss Driven Design Optimization for the Spallation Neutron Source (SNS) Ring

    International Nuclear Information System (INIS)

    Wei, J.

    1999-01-01

    This paper summarizes three-state design optimization for the Spallation Neutron Source (SNS) ring: linear machine design (lattice, aperture, injection, magnet field errors and misalignment), beam core manipulation (painting, space charge, instabilities, RF requirements), and beam halo consideration (collimation, envelope variation, e-p issues etc.)

  3. Optimum study designs.

    Science.gov (United States)

    Gu, C; Rao, D C

    2001-01-01

    Because simplistic designs will lead to prohibitively large sample sizes, the optimization of genetic study designs is critical for successfully mapping genes for complex diseases. Creative designs are necessary for detecting and amplifying the usually weak signals for complex traits. Two important outcomes of a study design--power and resolution--are implicitly tied together by the principle of uncertainty. Overemphasis on either one may lead to suboptimal designs. To achieve optimality for a particular study, therefore, practical measures such as cost-effectiveness must be used to strike a balance between power and resolution. In this light, the myriad of factors involved in study design can be checked for their effects on the ultimate outcomes, and the popular existing designs can be sorted into building blocks that may be useful for particular situations. It is hoped that imaginative construction of novel designs using such building blocks will lead to enhanced efficiency in finding genes for complex human traits.

  4. The advanced neutron source--designing to meet the needs of the user community

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1989-01-01

    The Advanced Neutron Source (ANS) is to be a multi-purpose neutron research center, constructed around a high-flux reactor now being designed at the Oak Ridge National Laboratory (ORNL). Its primary purpose is to place the United States in the forefront of neutron scattering in the twenty-first century. Other research programs include nuclear and fundamental physics, isotopes production, materials irradiation, and analytical chemistry. The Advanced Neutron Source will be a unique and invaluable research tool because of the unprecedented neutron flux available from the high intensity research reactor. But that reactor would be ineffective without world-class research facilities that allow the fullest utilization of the available neutrons. And, in turn, those research facilities will not produce new and exciting science without a broad population of users coming from all parts of the nation, and the world, placed in a simulating environment in which experiments can be effectively conducted, and in which scientific exchange is encouraged. This paper discusses the measures being taken to ensure that the design of the ANS focuses not only on the reactor, but on providing the experiment and user support facilities needed to allow its effective use. 5 refs., 4 figs

  5. Preliminary shielding analysis in support of the CSNS target station shutter neutron beam stop design

    Institute of Scientific and Technical Information of China (English)

    ZHANG Bin; CHEN Yi-Xue; WANG Wei-Jin; YANG Shou-Hai; WU Jun; YIN Wen; LIANG Tian-Jiao; JIA Xue-Jun

    2011-01-01

    The construction of China Spallation Neutron Source (CSNS) has been initiated in Dongguan,Guangdong, China.Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled Monte Carlo and multi-dimensional discrete ordinates method. The target of calculations is to optimize the neutron beamline shielding design to guarantee personal safety and minimize cost. Successful elimination of the primary ray effects via the two-dimensional uncollided flux and the first collision source methodology is also illustrated. Two-dimensional dose distribution is calculated. The dose at the end of the neutron beam line is less than 2.5μSv/h. The models have ensured that the doses received by the hall staff members are below the standard limit required.

  6. Design and investigations of a DD compact neutron generator in head radiotherapy

    International Nuclear Information System (INIS)

    Araujo, Wagner; Campos, Tarcisio Passos Ribeiro

    2013-01-01

    Neutron generators are device-based particle accelerators for producing neutrons through fusion reactions between hydrogen isotopes. Such devices may enable noninvasive treatments of head and neck tumors, which represent about one hundred twenty-nine thousand cases per year around the world. The present paper shows electromagnetic and nuclear simulations of a neutron generator coupled to collimator and evaluations of radiation dose in an analytical head phantom irradiated by the device. The results provide the generator design and the operation parameter in order to achieve prescribed tumor dose. Also, dose distribution in organs of head is presented, being suitable to surrounding brain tumors close to the skull. As conclusion, there is a visibility of neutron generator applied to brain tumor radiation therapy. (author)

  7. Neutronic data consistency analysis for lithium blanket and shield design

    International Nuclear Information System (INIS)

    Reupke, W.A.; Muir, D.W.

    1976-01-01

    Using a compact least-squares treatment we analyze the consistency of evaluated cross sections with calculated and measured tritium production in /sup n/Li and 7 Li detectors embedded in a 14-MeV neutron driven /sup n/LiD sphere. The tritium production experimental error matrix is evaluated and an initial reduced chi 2 of 3.0 is calculated. A perturbation calculation of the tritium production cross section sensitivities is performed with secondary neutron energy and angular distributions held constant. The cross section error matrix is evaluated by the external consistency of available cross section measurements. A statistical adjustment of the combined data yields a reduced chi 2 of 2.3 and represents a tenfold improvement in statistical likelihood. The improvement is achieved by a decrease in the 7 Li(n,xt) 14-MeV group cross section from 328 mb to 284 mb and an adjustment of the /sup n/Li data closer to calculated values. The uncertainty in the tritium breeding ratio in pure 7 LiD is reduced by one-fifth

  8. Preliminary neutron design of the flux flatter for silicon doping at the RA10

    International Nuclear Information System (INIS)

    Cintas, A.; Bazzana, S.

    2012-01-01

    The neutron transmutation doping of silicon (NTD) is one of the facilities under development for the RA10 project. In order to obtain high quality semiconductor, commercial requirements of NTD include achieving high axial and radial uniformity in the silicon targets. Axial uniformity is achieved locating a neutron screen around the Si ingot, obtaining a flat axial distribution of the dopant concentration. We present the neutron design of this screen, also known as flux flattener. MCNP5 was used to model the screen design. We have reached a satisfactory preliminary screen design after numerous iterations. The fluctuation in the axial distribution of the reaction capture rate ( 30 Si(n,γ) 31 Si) is under ≠1,5%, which is the required level by the semiconductor industry to accept the final product (author)

  9. Optimisation studies for a moderator on a pulsed neutron source

    International Nuclear Information System (INIS)

    Picton, D.J.; Ross, D.K.; Taylor, A.D.

    1982-01-01

    Having reviewed general aspects of moderator design for pulsed neutron sources, calculations are presented on a number of aspects of moderator optimization. Results of time-independent calculations on metal hydride moderators and a detailed method of evaluating moderated pulse intensities and time distributions, are given. Using computer codes, neutron cross-sections have been calculated from vibrational frequency distributions and time-dependent moderator calculations performed by Monte Carlo methods. The choice of an ambient moderator material and the optimum configuration of heterogeneous poisoning are examined and evaluations of liquid-nitrogen-cooled moderators are presented. Conclusions are drawn concerning the relative merits of cooled and poisoned moderators and an evaluation presented of solid methane at 20 K as a moderator for the production of cold neutrons. (U.K.)

  10. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  11. Uses of reactor neutrons for studying the microcomposition of materials

    International Nuclear Information System (INIS)

    Jervis, R.E.

    1993-01-01

    Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

  12. Neutron crystallographic studies of amino acids and nucleic acids

    International Nuclear Information System (INIS)

    Kashiwagi, Tatsuki

    2014-01-01

    Neutron crystallographic studies of two representative umami materials were executed utilizing iBLX at MLF/J-PARC. The results of them will be summarized in this report. At first, structure analysis of the alpha form crystal of L-glutamic acid was performed in order to assess the usefulness of neutron crystallography at iBIX to our company's R and D. Neutron crystal structure of it was successfully determined at 0.6 A resolution. All hydrogen atoms were clearly observed. Next, the mixed crystal of disodium Inosine-5'-phosphate (IMP · 2Na) and disodium Guanosine-5'-phosphate (GMP · 2Na) was analyzed by neutron crystallography. Neutron crystal structure of the mixed crystal of IMP and GMP (IM/GMP rate = 1.7) was successfully determined at 0.8 A resolution. In the neutron crystal structure of the mixed crystal, the hydrogen atom bonded to the C2 atom of purine base in IMP and the nitrogen atom bonded to the C2 atom of purine base in GMP were clearly observed in the nuclear density map, structurally demonstrating that this crystal is the mixed crystal. (author)

  13. Study and development of new dosemeters for thermal neutrons

    International Nuclear Information System (INIS)

    Urena N, F.

    1998-01-01

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  14. Studies on biological macromolecules by neutron inelastic scattering

    International Nuclear Information System (INIS)

    Fujiwara, Satoru; Nakagawa, Hiroshi

    2013-01-01

    Neutron inelastic scattering techniques, including quasielastic and elastic incoherent neutron scattering, provide unique tools to directly measure the protein dynamics at a picosecond time scale. Since the protein dynamics at this time scale is indispensable to the protein functions, elucidation of the protein dynamics is indispensable for ultimate understanding of the protein functions. There are two complementary directions of the protein dynamics studies: one is to explore the physical basis of the protein dynamics using 'model' proteins, and the other is more biology-oriented. Examples of the studies on the protein dynamics with neutron inelastic scattering are described. The examples of the studies in the former direction include the studies on the dynamical transitions of the proteins, the relationship between the protein dynamics and the hydration water dynamics, and combined analysis of the protein dynamics with molecular dynamics simulation. The examples of the studies in the latter direction include the elastic incoherent and quasielastic neutrons scattering studies of actin. Future prospects of the studies on the protein dynamics with neutron scattering are briefly described. (author)

  15. Study of the production of neutron-rich isotope beams issuing from fissions induced by fast neutrons; Etude de la production de faisceaux riches en neutrons par fission induite par neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Ch

    2000-09-15

    This work is a contribution to the PARRNe project (production of radioactive neutron-rich isotopes). This project is based on the fission fragments coming from the fission of 238-uranium induced by fast neutrons. The fast neutron flux is produced by the collisions of deutons in a converter. Thick targets of uranium carbide and liquid uranium targets have been designed in order to allow a quick release of fission fragments. A device, able to trap on a cryogenic thimble rare gas released by the target, has allowed the production of radioactive nuclei whose half-life is about 1 second. This installation has been settled to different deuton accelerators in the framework of the European collaboration SPIRAL-2. A calibration experiment has proved the feasibility of fixing an ISOL-type isotope separator to a 15 MV tandem accelerator, this installation can provide 500 nA deutons beams whose energy is 26 MeV and be a valuable tool for studying fast-neutron induced fission. Zinc, krypton, rubidium, cadmium, iodine, xenon and cesium beams have been produced in this installation. The most intense beams reach 10000 nuclei by micro-coulomb for 26 MeV deutons. An extra gain of 2 magnitude orders can be obtained by using a more specific ion source and by increasing the thickness of the target. Another extra gain of 2 magnitude orders involves 100 MeV deutons.

  16. Neutron activation analysis: Modelling studies to improve the neutron flux of Americium–Beryllium source

    Directory of Open Access Journals (Sweden)

    Abdessamad Didi

    2017-06-01

    Full Text Available Americium–beryllium (Am-Be; n, γ is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci, yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

  17. Design and optimization of a beam shaping assembly for BNCT based on D-T neutron generator and dose evaluation using a simulated head phantom.

    Science.gov (United States)

    Rasouli, Fatemeh S; Masoudi, S Farhad

    2012-12-01

    A feasibility study was conducted to design a beam shaping assembly for BNCT based on D-T neutron generator. The optimization of this configuration has been realized in different steps. This proposed system consists of metallic uranium as neutron multiplier, TiF(3) and Al(2)O(3) as moderators, Pb as reflector, Ni as shield and Li-Poly as collimator to guide neutrons toward the patient position. The in-air parameters recommended by IAEA were assessed for this proposed configuration without using any filters which enables us to have a high epithermal neutron flux at the beam port. Also a simulated Snyder head phantom was used to evaluate dose profiles due to the irradiation of designed beam. The dose evaluation results and depth-dose curves show that the neutron beam designed in this work is effective for deep-seated brain tumor treatments even with D-T neutron generator with a neutron yield of 2.4×10(12) n/s. The Monte Carlo Code MCNP-4C is used in order to perform these calculations. Copyright © 2012 Elsevier Ltd. All rights reserved.

  18. The design of a position-sensitive thermal-neutron detector

    International Nuclear Information System (INIS)

    Zhang Yi; Chen Ziyu; Shen Ji

    2007-01-01

    We design a type of position-sensitive thermal-neutron detector. The design is based on the nuclear reaction 10 B(n, α) 7 Li, and solid boron-10 is used as the target material while the alpha and lithium-7 particles from the reaction are caught as the source of position information of the original neutrons. With the help of MCNP software, we simulate the distribution of alpha particles in the boron target, which leads to the optimal thickness of target, physical efficiency and position resolution. (authors)

  19. Some preliminary design considerations for the ANS [Advanced Neutron Source] reactor cold source

    International Nuclear Information System (INIS)

    Henderson, D.L.

    1988-01-01

    Two areas concerned with the design of the Advanced Neutron Source (ANS) cold source have been investigated by simple one-dimensional calculations. The gain factors computed for a possible liquid nitrogen-15 cold source moderator are considerably below those computed for the much colder liquid deuterium moderator, as is reasonable considering the difference in moderator temperature. Nevertheless, nitrogen-15 does represent a viable option should safety related issues prohibit the use of deuterium as a moderating material. The slab geometry calculations have indicated that reflection of neutrons may be the dominant moderating mechanism and should be a consideration in the design of the cold source. 9 refs., 2 figs

  20. Target station design for a 1 MW pulsed spallation neutron source

    International Nuclear Information System (INIS)

    Russell, G.J.; Baker, G.D.; Brewton, R.J.

    1993-01-01

    Target stations are vital components of the 1 MW, next generation spallation neutron source proposed for LANSCE. By and large, target stations design determines the overall performance of the facility. Many traditional concepts will probably have to be rethought, and many new concepts will have to be put forward to meet the 1 MW challenge. This article gives a brief overview of the proposed neutron spallation source from the target station viewpoint, as well as the general philosophy adopted for the design of the LANSCE-II target stations. Some of the saliant concepts and features envisioned for LANSCE-II are briefly described

  1. Design of startup neutron detector handling mechanism instrumentation

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

    2010-01-01

    In PFBR, to monitor the reactor during first fuel loading and low power operation, special provision is made in the central fuel subassembly to accommodate the neutron detectors. During fuel handling operations, these detectors need to be lifted up to facilitate plug rotation. These detectors are also need to be lifted from the core to save their life, during intermediate and high power operations. Towards this, a mobile assembly containing these detectors is made with lowering and retracting provision. To control this operation, constant speed motor, torque limiter, proximity switch, wire drawn potentiometer, magnetic reed switches are provided. To ensure a smooth and safe handling of this assembly, control logic with necessary interlocks is developed. (author)

  2. A time-of-flight neutron reflectometer for surface and interfacial studies

    International Nuclear Information System (INIS)

    Penfold, J.; Ward, R.C.; Williams, W.G.

    1987-03-01

    A time-of-flight neutron reflectometer constructed for surface and interfacial studies, and installed at the ISIS pulsed neutron source, is described. One of its important design features is its inclined incident beam, since this allows both liquid and solid surface phenomena to be investigated. Measurements are presented to show the performance of the instrument, and new representative results, which include studies of liquid surfaces, Langmuir-Blodgett films, and thin film multilayers, are included as illustrations of the scientific potential of the method. (author)

  3. Neutron/photon/electron shielding study for a laser-fusion facility

    International Nuclear Information System (INIS)

    Thompson, W.L.

    1977-01-01

    A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

  4. A multiwire proportional chamber for precision studies of neutron β decay angular correlations

    International Nuclear Information System (INIS)

    Ito, T.M.; Carr, R.; Filippone, B.W.; Martin, J.W.; Plaster, B.; Rybka, G.; Yuan, J.

    2007-01-01

    A new multiwire proportional chamber (MWPC) was designed and constructed for precision studies of neutron β decay angular correlations. Its design has several novel features, including the use of low pressure neopentane as the MWPC gas and an entrance window made of thin Mylar sheet reinforced with Kevlar fibers. In the initial off-line performance tests, the gas gain of neopentane and the position resolution were studied

  5. Design of emergency shutdown system for the Tehran Research Reactor; Part I: Neutronics investigation

    International Nuclear Information System (INIS)

    Safarinia, M.; Faghihi, F.; Mirvakili, S.M.; Fakhraei, A.

    2017-01-01

    Highlights: • An emergency shutdown system for the TRR is carried out based on a heavy water tank. • The performance of the heavy water tank are carried out based on “first and equilibrium cores”. • Heavy water discharging flow rate is also studied in the current research. • Thermal flux in the radioisotope channel with and without the heavy water tank are studied. • A core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). - Abstract: In this paper, a neutronics design of the secondary (i.e., emergency) shutdown system for the Tehran Research Reactor (TRR) is carried out based on a heavy water tank design. The heavy water tank in a cylindrical shape is around the core, and calculations for the optimized radius and height of the tank are performed. The performance of the heavy water tank calculations are carried out based on two types of fuel loading, which are called the “first and equilibrium cores” of the TRR. For both cases, neutronics and standard safety analysis are taken into account, benchmarked, and described herein. Heavy water discharging flow rate is also studied in the current research, and the results are compared with the IAEA criteria. Moreover, thermal flux in the radioisotope channel with and without the heavy water tank (as the reflector) are studied herein. Specifically, a core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). Based on our optimization, the 5 × 5 fuel assembly, which is called “B configuration,” has better performance and efficiency than that of the other described layouts.

  6. Preliminary neutronic study on Pu-based OTTO cycle pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Setiadipura, Topan; Zuhair [National Nuclear Energy Agency of Indonesia (BATAN), Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Irwanto, Dwi [Bandung Institute of Technology (ITB), Bandung (Indonesia). Nuclear Physics and Biophysics Research Group

    2017-12-15

    The neutron physics characteristic of Pebble Bed Reactor (PBR) allows a better incineration of plutonium (Pu). An optimized design of simple PBR might give a symbiotic solution of providing a safe energy source, effective fuel utilization shown by a higher burnup value, and incineration of Pu stockpiles. This study perform a preliminary neutronic design study of a 200 MWt Once Through Then Out (OTTO) cycle PBR with Pu-based fuel. The safety criteria of the design were represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. In this preliminary phase, the parametric survey is limited to the heavy metal (HM) loading per pebble and the average axial speed of the fuel. An optimum high burnup of 419.7 MWd/kg-HM was achieved in this study. This optimum design uses a HM loading of 2.5 g/pebble with average axial fuel velocity 0.5 cm/day.

  7. Design of analytical instrumentation with D-T sealed neutron generators

    International Nuclear Information System (INIS)

    Qiao Yahua; Wu Jizong; Zheng Weiming; Liu Quanwei; Zhang Min

    2008-01-01

    Analytical instrumentation with D-T sealed neutron generators source activation, The 14 MeV D-T sealed neutron tube with 10 9 n · s -1 neutron yield is used as generator source. The optimal structure of moderator and shield was achieved by MC computing.The instrumentation's configuration is showed. The instrumentation is made up of the SMY-DT50.8-2.1 sealed neutron tube and the high-voltage power supply system, which center is the sealed neutron generators. 6 cm Pb and 20 cm polythene is chosen as moderator, Pb, polythene and 10 cm boron-PE was chosen as shield .The sample box is far the source from 9 cm, the measurement system were made up of HPGe detector and the sample transforming system. After moderator and shield, the thermal neutron fluence rate at the point of sample is 0.93 × 10 6 n · s -1 cm -2 , which is accorded with design demand, and the laboratory and surroundings reaches the safety standard of the dose levels. (authors)

  8. Detailed design of neutron guide tubes at the upgraded JRR-3, (1)

    International Nuclear Information System (INIS)

    Harami, Taikan; Umemura, Mutsumi; Ebisawa, Tohru.

    1985-07-01

    JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. Two guide tubes for thermal neutron and three for cold will be installed in the reactor to transport thermal and cold neutrons from the reactor hall to the experiment hall. This describes the neutron guide tube transmission analysis program, NEUGT, which was developed to assess the design of the neutron guide tubes. The input data plotting program, PLOPINE and the output data plotting program, NEUPLOT are presented in the appendix. The NEUGT program not only calculates a neutron transmission and neutron spectra, assuming the Maxwellian spectra at the entrance of a guide tube, but also analyses the effect of abutment errors. This reports the description and the input data manual of the program in the text. Examples of analysis are given in the appendixes. The program is written in the FORTRAN 77 language for FACOM 380. (author)

  9. Design of an electron-accelerator-driven compact neutron source for non-destructive assay

    Science.gov (United States)

    Murata, A.; Ikeda, S.; Hayashizaki, N.

    2017-09-01

    The threat of nuclear and radiological terrorism remains one of the greatest challenges to international security, and the threat is constantly evolving. In order to prevent nuclear terrorism, it is important to avoid unlawful import of nuclear materials, such as uranium and plutonium. Development of technologies for non-destructive measurement, detection and recognition of nuclear materials is essential for control at national borders. At Tokyo Institute of Technology, a compact neutron source system driven by an electron-accelerator has been designed for non-destructive assay (NDA). This system is composed of a combination of an S-band (2.856 GHz) RF-gun, a tungsten target to produce photons by bremsstrahlung, a beryllium target, which is suitable for use in generating neutrons because of the low threshold energy of photonuclear reactions, and a moderator to thermalize the fast neutrons. The advantage of this system can accelerate a short pulse beam with a pulse width less than 1 μs which is difficult to produce by neutron generators. The amounts of photons and neutron produced by electron beams were simulated using the Monte Carlo simulation code PHITS 2.82. When the RF-gun is operated with an average electron beam current of 0.1 mA, it is expected that the neutron intensities are 1.19 × 109 n/s and 9.94 × 109 n/s for incident electron beam energies of 5 MeV and 10 MeV, respectively.

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    International Nuclear Information System (INIS)

    Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

    2015-01-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  12. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

    2015-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  13. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, Kevin James [Univ. of California, Berkeley, CA (United States)

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

  14. Design status of an intense 14 MeV neutron source for cancer therapy

    CERN Document Server

    Yao, Z E; Cheng, S W; Jia, W B

    2002-01-01

    Design and development of an intense 14 MeV neutron source for cancer therapy is in progress at the Institute of Nuclear Research of Lanzhou University. The neutrons from the T(d,n) sup 4 He reaction are produced by bombarding a rotating titanium tritide target with a 40 mA deuteron beam at 600 keV. The designed neutron yield is 8x10 sup 1 sup 2 n/s and the maximum dose rate at a 100 cm source-to-skin distance is 25 cGy/min. The HV terminal, accelerating column and HV power supply are enclosed inside a stainless steel pressure vessel containing 6 atm SF sub 6 gas to provide the electrical insulation.

  15. The design of a proton recoil telescope for 14 MeV neutron spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Hawkes, N.P.; Bond, D.S.; Croft, S.; Jarvis, O.N. E-mail: onj@jet.uk; Sherwood, A.C

    2002-01-01

    As part of the design effort for a 14 MeV neutron spectrometer for the Joint European Torus (JET), computer codes were developed to calculate the response of a proton recoil telescope comprising a proton radiator film mounted in front of a proton detector. The codes were used to optimise the geometrical configuration in terms of efficiency and resolution, bearing in mind the constraints imposed by the proposed application as a JET neutron diagnostic for the Deuterium-Tritium phase. A prototype instrument was built according to the optimised design, and tested with monoenergetic 14 MeV neutrons from the Harwell 500 keV Van de Graaff accelerator. The measured energy resolution and absolute efficiency were found to be in acceptable agreement with the calculations. Based on this work, a multi-radiator production version of the spectrometer has now been constructed and successfully deployed at JET.

  16. The design of a proton recoil telescope for 14 MeV neutron spectrometry

    International Nuclear Information System (INIS)

    Hawkes, N.P.; Bond, D.S.; Croft, S.; Jarvis, O.N.; Sherwood, A.C.

    2002-01-01

    As part of the design effort for a 14 MeV neutron spectrometer for the Joint European Torus (JET), computer codes were developed to calculate the response of a proton recoil telescope comprising a proton radiator film mounted in front of a proton detector. The codes were used to optimise the geometrical configuration in terms of efficiency and resolution, bearing in mind the constraints imposed by the proposed application as a JET neutron diagnostic for the Deuterium-Tritium phase. A prototype instrument was built according to the optimised design, and tested with monoenergetic 14 MeV neutrons from the Harwell 500 keV Van de Graaff accelerator. The measured energy resolution and absolute efficiency were found to be in acceptable agreement with the calculations. Based on this work, a multi-radiator production version of the spectrometer has now been constructed and successfully deployed at JET

  17. Study of high-j neutron excitations outside 136Xe

    Science.gov (United States)

    Talwar, R.; Kay, B. P.; Mitchell, A. J.; Adachi, S.; Entwisle, J. P.; Fujita, Y.; Gey, G.; Noji, S.; Ong, H. J.; Schiffer, J. P.; Tamii, A.

    2017-09-01

    The character of single-neutron excitations outside of N = 82 has been studied using nucleon transfer reactions in terms of the energy centroid of their strength as well as the fragmentation of this strength among the actual states of the nucleus. However, extending the systematic study of the N = 83 isotones to 137Xe has been challenging due to xenon being a gas at room temperature. Though several attempts have been made, a quantitative determination of the spectroscopic factors for the neutron 9/2- and 13/2+ excitations in 137Xe is still lacking. In the present work, we report on a study of the 136Xe(α,3He)137Xe reaction carried out at 100 MeV to probe the l = 5 , 9/2- and l = 6 , 13/2+ single-neutron excitations. The experimental technique and results will be presented discussing them in context of the evolution of these single-neutron excitations and the influence of the tensor interaction on the neutron single-particle states as the proton orbits are filling. This work has been supported by the U.S. Department of Energy, Office of Science, Office of Nuclear Physics, under Contract Number DE-AC02-06CH11357, the Australian Research Council Discovery Project 120104176, and the UK Science and Technology Facilities.

  18. Inelastic Neutron Scattering Study of Mn

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Y.; Sarachik, M.P.; Friedman, J.R.; Robinson, R.A.; Kelley, T.M.; Nakotte, H.; Christianson, A.C.; Trouw, F.; Aubin, S.M.J.; Hendrickson, D.N.

    1998-11-09

    The authors report zero-field inelastic neutron scattering experiments on a 14-gram deuterated sample of Mn{sub 12}-Acetate consisting of a large number of identical spin-10 magnetic clusters. Their resolution enables them to see a series of peaks corresponding to transitions between the anisotropy levels within the spin-10 manifold. A fit to the spin Hamiltonian H = {minus}DS{sub z}{sup 2} + {mu}{sub B}B{center_dot}g{center_dot}S-BS{sub z}{sup 4} + C(S{sub +}{sup 4} + S{sub {minus}}{sup 4}) yields an anisotropy constant D = (0.54 {+-} 0.02) K and a fourth-order diagonal anisotropy coefficient B = (1.2 {+-} 0.1) x 10{sup {minus}3}K. Unlike EPR measurements, their experiments do not require a magnetic field and yield parameters that do not require knowledge of the g-value.

  19. Neutron Inelastic Scattering Study of Liquid Argon

    Energy Technology Data Exchange (ETDEWEB)

    Skoeld, K; Rowe, J M; Ostrowski, G [Solid State Science Div., Argonne National Laboratory, Argonne, Illinois (US); Randolph, P D [Nuclear Technology Div., Idaho Nuclear Corporation, Idaho Falls, Idaho (US)

    1972-02-15

    The inelastic scattering functions for liquid argon have been measured at 85.2 K. The coherent scattering function was obtained from a measurement on pure A-36 and the incoherent function was derived from the result obtained from the A-36 sample and the result obtained from a mixture of A-36 and A-40 for which the scattering is predominantly incoherent. The data, which are presented as smooth scattering functions at constant values of the wave vector transfer in the range 10 - 44/nm, are corrected for multiple scattering contributions and for resolution effects. Such corrections are shown to be essential in the derivation of reliable scattering functions from neutron scattering data. The incoherent data are compared to recent molecular dynamics results and the mean square displacement as a function of time is derived. The coherent data are compared to molecular dynamics results and also, briefly, to some recent theoretical models

  20. Designing satisfaction studies

    DEFF Research Database (Denmark)

    Kristensen, Kai; Eskildsen, Jacob Kjær

    2007-01-01

    In the effect sampling method, presentation of researcher, the intro text, the order of questions in the questionnaire along with the number of categories in the rating scale is tested in relation to the design of satisfaction studies. Based on the analyses specific recommendations for designing...... satisfaction studies are given....