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Sample records for responsive neutron generator

  1. Investigation of Response of Several Neutron Surveymeters by a DT Neutron Generator

    International Nuclear Information System (INIS)

    Kim, Sang In; Jang, In Su; Kim, Jang Lyul; Lee, Jung IL; Kim, Bong Hwan

    2012-01-01

    Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

  2. High energy neutron generator

    International Nuclear Information System (INIS)

    Barjon, R.; Breynat, G.

    1987-01-01

    This patent describes a generator of fast neutrons only slightly contaminated by neutrons of energy less than 15 MeV, comprising a source of charged particles of energy equal to at least 15 MeV, a target made of lithium deuteride, and means for cooling the target. The target comprises at least two elements placed in series in the path of the charged particles and separated from each other, the thickness of each of the elements being selected as a function of the average energy of the charged particles emitted from the source and the energy of the fast neutrons to be generated such that neutrons of energy equal to at least 15 MeV are emitted in the forward direction in response to the bombardment of the target from behind by the charged particles. The target cooling means comprises means for circulating between and around the elements a gas which does not chemically react with lithium deuteride

  3. Pulsed neutron generator

    International Nuclear Information System (INIS)

    Bespalov, D.F.; Bykovskii, Yu.A.; Vergun, I.I.; Kozlovskii, K.I.; Kozyrev, Yu.P.; Leonov, R.K.; Simagin, B.I.; Tsybin, A.S.; Shikanov, A.Ie.

    1986-03-01

    The paper describes a new device for generating pulsed neutron fields, utilized in nuclear geophysics for carrying out pulsed neutron logging and activation analysis under field conditions. The invention employs a sealed-off neutron tube with a laser ion source which increases neutron yield to the level of 10 neutrons per second or higher. 2 refs., 1 fig

  4. Neutron generator control system

    International Nuclear Information System (INIS)

    Peelman, H.E.; Bridges, J.R.

    1981-01-01

    A method is described of controlling the neutron output of a neutron generator tube used in neutron well logging. The system operates by monitoring the target beam current and comparing a function of this current with a reference voltage level to develop a control signal used in a series regulator to control the replenisher current of the neutron generator tube. (U.K.)

  5. Nanosecond neutron generator

    International Nuclear Information System (INIS)

    Lobov, S.I.; Pavlovskaya, N.G.; Pukhov, S.P.

    1991-01-01

    High-voltage nanosecond neutron generator for obtaining neutrons in D-T reaction is described. Yield of 6x10 6 neutron/pulse was generated in a sealed gas-filled diode with a target on the cathode by accelerating pulse voltage of approximately 0.5 MV and length at half-height of 0.5 ns and deuterium pressure of 6x10 -2 Torr. Ways of increasing neutron yield and possibilities of creating generators of nanosecond neutron pulses with great service life are considered

  6. Cylindrical neutron generator

    Science.gov (United States)

    Leung, Ka-Ngo [Hercules, CA

    2008-04-22

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  7. Compact neutron generator

    Science.gov (United States)

    Leung, Ka-Ngo; Lou, Tak Pui

    2005-03-22

    A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

  8. The intense neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1966-07-01

    The presentation discusses both the economic and research contexts that would be served by producing neutrons in gram quantities at high intensities by electrical means without uranium-235. The revenue from producing radioisotopes is attractive. The array of techniques introduced by the multipurpose 65 megawatt Intense Neutron Generator project includes liquid metal cooling, superconducting magnets for beam bending and focussing, super-conductors for low-loss high-power radiofrequency systems, efficient devices for producing radiofrequency power, plasma physics developments for producing and accelerating hydrogen, ions at high intensity that are still far out from established practice, a multimegawatt high voltage D.C. generating machine that could have several applications. The research fields served relate principally to materials science through neutron-phonon and other quantum interactions as well as through neutron diffraction. Nuclear physics is served through {mu}-, {pi}- and K-meson production. Isotope production enters many fields of applied research. (author)

  9. The intense neutron generator

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1966-01-01

    The presentation discusses both the economic and research contexts that would be served by producing neutrons in gram quantities at high intensities by electrical means without uranium-235. The revenue from producing radioisotopes is attractive. The array of techniques introduced by the multipurpose 65 megawatt Intense Neutron Generator project includes liquid metal cooling, superconducting magnets for beam bending and focussing, super-conductors for low-loss high-power radiofrequency systems, efficient devices for producing radiofrequency power, plasma physics developments for producing and accelerating hydrogen, ions at high intensity that are still far out from established practice, a multimegawatt high voltage D.C. generating machine that could have several applications. The research fields served relate principally to materials science through neutron-phonon and other quantum interactions as well as through neutron diffraction. Nuclear physics is served through μ-, π- and K-meson production. Isotope production enters many fields of applied research. (author)

  10. Spherical neutron generator

    Science.gov (United States)

    Leung, Ka-Ngo

    2006-11-21

    A spherical neutron generator is formed with a small spherical target and a spherical shell RF-driven plasma ion source surrounding the target. A deuterium (or deuterium and tritium) ion plasma is produced by RF excitation in the plasma ion source using an RF antenna. The plasma generation region is a spherical shell between an outer chamber and an inner extraction electrode. A spherical neutron generating target is at the center of the chamber and is biased negatively with respect to the extraction electrode which contains many holes. Ions passing through the holes in the extraction electrode are focused onto the target which produces neutrons by D-D or D-T reactions.

  11. Neutron generation in lightning bolts

    International Nuclear Information System (INIS)

    Shah, G.N.; Razdan, H.; Bhat, C.L.; Ali, Q.M.

    1985-01-01

    To ascertain neutron generation in lightning bolts, the authors have searched for neutrons from individual lightning strokes, for a time-interval comparable with the duration of the lightning stroke. 10 7 -10 10 neutrons per stroke were found, thus providing the first experimental evidence that neutrons are generated in lightning discharges. (U.K.)

  12. Evaluation of area monitor response for neutrons in radiation field generated by a 15 MV clinic accelerator

    International Nuclear Information System (INIS)

    Salgado, Ana Paula

    2011-01-01

    The clinical importance and usage of linear accelerators in cancer treatment increased significantly in the last years. Coupled with this growth came the concern about the use of accelerators with energies over to 10 MeV which produce therapeutic beam contaminated with neutrons generated when high-energy photons interact with high-atomic-number materials such as tungsten and lead present in the accelerator itself. At these facilities, measurements of the ambient dose equivalent for neutrons present difficulties owing to the existence of a mixed radiation field and possible electromagnetic interference near the accelerator. The Neutron Laboratory of the IRD - Brazilian Institute for Radioprotection and Dosimetry, aiming to evaluate the survey meters performance at these facilities, initiated studies of instrumentation response in the presence of different neutron spectra. Neutrons sources with average energies ranging from 0.55 to 4.2 MeV, four different survey meters and one ionization chamber to obtain the ratio between the dose due to neutrons and gamma radiation were used in this work. The evaluation of these measurements, performed in a 15 MV linear accelerator room is presented. This work presents results that demonstrate the complexity and care needed to make neutrons measurements in radiotherapy treatment rooms containing high energy clinical accelerators. (author)

  13. Short pulse neutron generator

    Science.gov (United States)

    Elizondo-Decanini, Juan M.

    2016-08-02

    Short pulse neutron generators are described herein. In a general embodiment, the short pulse neutron generator includes a Blumlein structure. The Blumlein structure includes a first conductive plate, a second conductive plate, a third conductive plate, at least one of an inductor or a resistor, a switch, and a dielectric material. The first conductive plate is positioned relative to the second conductive plate such that a gap separates these plates. A vacuum chamber is positioned in the gap, and an ion source is positioned to emit ions in the vacuum chamber. The third conductive plate is electrically grounded, and the switch is operable to electrically connect and disconnect the second conductive plate and the third conductive plate. The at least one of the resistor or the inductor is coupled to the first conductive plate and the second conductive plate.

  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. Pulsed neutron generator for logging

    International Nuclear Information System (INIS)

    Thibideau, F.D.

    1977-01-01

    A pulsed neutron generator for uranium logging is described. This generator is one component of a prototype uranium logging probe which is being developed by SLA to detect, and assay, uranium by borehole logging. The logging method is based on the measurement of epithermal neutrons resulting from the prompt fissioning of uranium from a pulsed source of 17.6 MeV neutrons. An objective of the prototype probe was that its diameter not exceed 2.75 inches, which would allow its use in conventional rotary drill holes of 4.75-inch diameter. This restriction limited the generator to a maximum 2.375-inch diameter. The performance requirements for the neutron generator specified that it operate with a nominal output of 5 x 10 6 neutrons/pulse at up to 100 pulses/second for a one-hour period. The development of a neutron generator meeting the preliminary design goals was completed and two prototype models were delivered to SLA. These two generators have been used by SLA to log a number of boreholes in field evaluation of the probe. The results of the field evaluations have led to the recommendation of several changes to improve the probe's operation. Some of these changes will require additional development effort on the neutron generator. It is expected that this work will be performed during 1977. The design and operation of the first prototype neutron generators is described

  16. A compact mobile neutron generator

    International Nuclear Information System (INIS)

    Zhou Changgeng; Li Yan; Hu Yonghong; Lou Benchao; Wu Chunlei

    2007-06-01

    Through fitting the high voltage terminal from introducing overseas and pulse system et al. from oneself developing together, a compact mobile neutron generator is established. The length and weight of this neutron generator are 2 500 mm and less than 1 t, respectively. It can be expediently moved to the location which is required by experimental people. It is consisted of RF ion source, acceleration tube, high voltage generator, focus device, microsecond pulse system, gas leak system, control system, vacuum system and experimental target. It can produce 150 μA continuous deuterium ion beam current, also can produce the pulse deuterium ion beam current. The pulse widths are 10-100 μs and frequencies 10 Hz, 1 000 Hz, 10 000 Hz. The D-T neutron yields of the neutron generator may arrive 1.5 x 10 10 s -1 . The working principle and the structure of the main parts of this neutron generator are described. (authors)

  17. Neutron generator tube ion source control

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

    A system is claimed for controlling the output of a neutron generator tube of the deuterium-tritium accelerator type and having an ion source to produce sharply defined pulses of neutrons for well logging use. It comprises: means for inputting a relatively low voltage input control pulse having a leading edge and a trailing edge; means, responsive to the input control pulse, for producing a relatively high voltage ion source voltage pulse after receipt of the input pulse; and means, responsive to the input control pulse, for quenching, after receipt of the input pulse, the ion source control pulse, thereby providing a sharply time defined neutron output from the generator tube

  18. Neutron generators at Purnima Lab

    International Nuclear Information System (INIS)

    Patel, Tarun; Sinha, Amar

    2015-01-01

    Neutron sources are in a great demand in many area like research, nuclear waste management, industrial process control, medical and also security. Major sources of neutrons are nuclear reactors, radioisotopes and accelerator based neutron generators. For many field applications, reactors cannot be used due to its large size, complicated system, high cost and also safety issues. Radioisotopes like Pu-Be, Am-Be, Cf, are extensively used for many industrial applications. But they are limited in their use due to their low source strength and also handling difficulties due to radioactivity. They are also not suitable for pulsed neutron applications. In contrast, compact size, pulsed operation, on/off operation etc.of accelerator based neutron generators make them very popular for many applications. Particle accelerators based on different types of neutron generators have been developed around the world. Among these deuteron accelerator based D-D and D-T neutron generators are widely used as they produce mono-energetic fast neutrons and in particular high yield of D-T neutron can be obtained with less than 300 KV of accelerating voltage

  19. Laser neutron generator

    International Nuclear Information System (INIS)

    Anan'in, O.B.; Bespalov, D.F.; Bykovskii, Yu.A.; Kozyrev, Yu.P.; Mints, A.Z.; Riabov, E.V.; Tsybin, A.S.; Cherkasov, Yu.; Shikanov, A.E.

    1986-01-01

    Information is presented concerning devices for producing intense neutrons flows, and may be utilized in nuclear geophysics for carrying out pulsed neutron logging of wells, in studies of the critical characteristics of nuclear reactors, for activation analysis, radiation therapy, defectoscopy, and so on

  20. Generation of organic scintillators response function for fast neutrons using the Monte Carlo method

    International Nuclear Information System (INIS)

    Mazzaro, A.C.

    1979-01-01

    A computer program (DALP) in Fortran-4-G language, has been developed using the Monte Carlo method to simulate the experimental techniques leading to the distribution of pulse heights due to monoenergetic neutrons reaching an organic scintillator. The calculation of the pulse height distribution has been done for two different systems: 1) Monoenergetic neutrons from a punctual source reaching the flat face of a cylindrical organic scintillator; 2) Environmental monoenergetic neutrons randomly reaching either the flat or curved face of the cylindrical organic scintillator. The computer program has been developed in order to be applied to the NE-213 liquid organic scintillator, but can be easily adapted to any other kind of organic scintillator. With this program one can determine the pulse height distribution for neutron energies ranging from 15 KeV to 10 MeV. (Author) [pt

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

    International Nuclear Information System (INIS)

    Rochau, G.E.

    1980-01-01

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

  2. High efficiency focus neutron generator

    Science.gov (United States)

    Sadeghi, H.; Amrollahi, R.; Zare, M.; Fazelpour, S.

    2017-12-01

    In the present paper, the new idea to increase the neutron yield of plasma focus devices is investigated and the results are presented. Based on many studies, more than 90% of neutrons in plasma focus devices were produced by beam target interactions and only 10% of them were due to thermonuclear reactions. While propounding the new idea, the number of collisions between deuteron ions and deuterium gas atoms were increased remarkably well. The COMSOL Multiphysics 5.2 was used to study the given idea in the known 28 plasma focus devices. In this circumstance, the neutron yield of this system was also obtained and reported. Finally, it was found that in the ENEA device with 1 Hz working frequency, 1.1 × 109 and 1.1 × 1011 neutrons per second were produced by D-D and D-T reactions, respectively. In addition, in the NX2 device with 16 Hz working frequency, 1.34 × 1010 and 1.34 × 1012 neutrons per second were produced by D-D and D-T reactions, respectively. The results show that with regards to the sizes and energy of these devices, they can be used as the efficient neutron generators.

  3. The Sealed Tube Neutron Generator

    International Nuclear Information System (INIS)

    Tunnell, L.N.; Beyerle, A.; Durkee, R.; Headley, G.; Hurley, P.

    1992-01-01

    A Sealed Tube Neutron Generator (STNG) has been designed and tested at Special Technologies Laboratories (STL) in Santa Barbara, California. Unlike similar tubes that have been used for years in other applications, e.g., by the oil well logging industry, the present device was designed primarily to be part of the Associated Particle Imaging (API) system. Consequently, the size and quality of the neutron spot produced by the STNG is of primary importance. Results from initial measurements indicate that performance goals are satisfied

  4. Compact ion source neutron generator

    Science.gov (United States)

    Schenkel, Thomas; Persaud, Arun; Kapadia, Rehan; Javey, Ali; Chang-Hasnain, Constance; Rangelow, Ivo; Kwan, Joe

    2015-10-13

    A neutron generator includes a conductive substrate comprising a plurality of conductive nanostructures with free-standing tips and a source of an atomic species to introduce the atomic species in proximity to the free-standing tips. A target placed apart from the substrate is voltage biased relative to the substrate to ionize and accelerate the ionized atomic species toward the target. The target includes an element capable of a nuclear fusion reaction with the ionized atomic species to produce a one or more neutrons as a reaction by-product.

  5. Source characterization of Purnima Neutron Generator (PNG)

    International Nuclear Information System (INIS)

    Bishnoi, Saroj; Patel, T.; Paul, Ram K.; Sarkar, P.S.; Adhikari, P.S.; Sinha, Amar

    2011-01-01

    The use of 14.1 MeV neutron generators for the applications such as elemental analysis, Accelerated Driven System (ADS) study, fast neutron radiography requires the characterization of neutron source i.e neutron yield (emission rate in n/sec), neutron dose, beam spot size and energy spectrum. In this paper, a series of experiments carried out to characterize this neutron source. The neutron source has been quantified with neutron emission rate, neutron dose at various source strength and beam spot size at target position

  6. Neutron generator ion source pulser

    International Nuclear Information System (INIS)

    Peelman, H.E.

    1987-01-01

    This patent describes, for use with a pulsed neutron generator in a logging tool lowered in a borehole, a pulsed high voltage source having an output terminal adapted to be connected to pulse neutron generator. The power supply comprises: (a) high voltage supply means; (b) field effect transistor means comprising at least a pair of field effect transistors serially connected between the high voltage supply means and ground; (c) an output terminal between the two transistors of the field effect transistor means, the output terminal adapted to be connected by a conductor to provide pulsed high voltage to a neutron generator; (d) control pulse forming means connected to the gates of the respective two transistors, the pulse forming means forming control pulses selectively switching the transistors off and on in timed sequence to thereby connect the output terminal to the high voltage supply means, and (e) diode means connected to the gates of the transistors to limit gate voltage for operation of the transistors

  7. 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.

  8. Evaluation of area monitor response for neutrons in radiation field generated by a 15 MV clinic accelerator; Avaliacao da resposta dos monitores de area para neutrons em campo de radiacao gerado por um acelerador clinico de 15 MV

    Energy Technology Data Exchange (ETDEWEB)

    Salgado, Ana Paula

    2011-07-01

    The clinical importance and usage of linear accelerators in cancer treatment increased significantly in the last years. Coupled with this growth came the concern about the use of accelerators with energies over to 10 MeV which produce therapeutic beam contaminated with neutrons generated when high-energy photons interact with high-atomic-number materials such as tungsten and lead present in the accelerator itself. At these facilities, measurements of the ambient dose equivalent for neutrons present difficulties owing to the existence of a mixed radiation field and possible electromagnetic interference near the accelerator. The Neutron Laboratory of the IRD - Brazilian Institute for Radioprotection and Dosimetry, aiming to evaluate the survey meters performance at these facilities, initiated studies of instrumentation response in the presence of different neutron spectra. Neutrons sources with average energies ranging from 0.55 to 4.2 MeV, four different survey meters and one ionization chamber to obtain the ratio between the dose due to neutrons and gamma radiation were used in this work. The evaluation of these measurements, performed in a 15 MV linear accelerator room is presented. This work presents results that demonstrate the complexity and care needed to make neutrons measurements in radiotherapy treatment rooms containing high energy clinical accelerators. (author)

  9. PNG-300 a nanosecond pulsed neutron generator

    International Nuclear Information System (INIS)

    Sztaricskai, T.; Vasvary, L.; Petoe, G.C.; Devkin, B.V.

    1985-01-01

    The design and operation of a nanosecond-pulse neutron generator is reported. It was constructed for the measurement of prompt neutron and gamma radiation in experimental studies of fast neutron reactions by time of flight techniques. The acceleration voltage is 300 kV and the total resolution of the generator-neutron spectrometer system is 2 ns. The ion-optical system, the vacuum system and the control of the neutron generator is described in detail. The equipment was used for prompt neutron and gamma radiation induced in construction materials. (R.P.)

  10. The Canadian intense neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Tunnicliffe, P R

    1967-07-01

    Atomic Energy of Canada Ltd. has proposed construction of an Intense Neutron-Generator. The generator would produce uniquely-intense beams of thermal neutrons for solid-state and low-energy nuclear studies and would yield significant quantities of radioisotopes of both research and commercial value; it would also produce copious sources of mesons and energetic nucleons for use in intermediate-energy nuclear physics and in nuclear-structure studies. The primary neutron source of 10{sup 19}/sec would be generated by bombarding a heavy-element target with a continuous beam of 65 mA of 1 GeV protons. The target of circulating and cooled Pb-Bi eutectic would be surrounded by a tank of heavy water moderator yielding a maximum useful flux of 10{sup 16} thermal neutrons/cm{sup 2}/sec in the region where neutron beams can be extracted. This high-energy spallation process for producing neutrons is nearly four times more efficient in producing neutrons per unit of thermal energy released in the neutron source compared with a fission reactor. Nevertheless, if energy costs for producing the 65 MW proton beam are to be within reason, the machine producing the beam must be efficient. A D.C. machine is in principle ideal but practical achievement of 1 GV is not likely within the time desired. An accelerator where the protons gain energy from radio-frequency fields is the most likely prospect. We have selected a linear accelerator as our reference design and detailed theoretical and experimental studies are in progress. The machine is based on the Los Alamos Meson Physics Facility design reoptimized for continuous rather than pulsed operation. It is approximately one mile long and is expected to achieve nearly 50 percent overall efficiency. There are two major portions, an 'Alvarez' Section operating at 200 MHz accelerating the beam to about 150 MeV, followed by a 'Waveguide' section operating at 800 MHz. Protons are initially injected by an 0.75 MV D.C. accelerator. The Alvarez

  11. Optimization of the beam shaping assembly in the D-D neutron generators-based BNCT using the response matrix method.

    Science.gov (United States)

    Kasesaz, Y; Khalafi, H; Rahmani, F

    2013-12-01

    Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45 MeV neutrons that are produced in the D-D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70 cm Fluental as a moderator, 30 cm Pb as a reflector, 2mm (6)Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method. Copyright © 2013 Elsevier Ltd. All rights reserved.

  12. Experiment of Neutron Generation by Using 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; Park, Ji Young; Hwang, Yong Seok; Choi, H.D.

    2005-01-01

    Experiment of neutron generation was performed by using a prototype D-D neutron generator. The characteristics of D-D neutron generation in drive-in target was studied. The increment of neutron yield by increasing ion beam energy was investigated, too

  13. Compact neutron generators for environmental recovery applications

    International Nuclear Information System (INIS)

    Leung, K. N.; Firestone, R. B.; Lou, T. P.; Reijonen, J.; Vujic, J. Lj.

    2002-01-01

    New generations of compact neutron sources are being developed at the Lawrence Berkeley National Laboratory (LBNL). The D-D or D-T neutron generators can be used to perform precise elemental analysis by Prompt Gamma-Ray Activation Analysis (PGAA) in place of a nuclear reactor. The neutron generators will be composed of an ion source, from which a 1.5 A deuterium beam will be extracted and accelerated to about 150 keV onto a target loaded with deuterium. Based on the D-D nuclear reaction, the neutron generator will yield approximately 10 12 n/s (10 14 n/s for D-T reaction). With this neutron output, thermal and cold neutron fluxes of 10 7 n/s cm 2 and 6 x 10 6 n/s cm 2 have been estimated using neutron moderators designed by the neutron transport simulation code MCNP. (author)

  14. The secondary neutron sources for generation of particular neutron fluxes

    International Nuclear Information System (INIS)

    Tracz, G.

    2007-07-01

    The foregoing paper presents the doctor's thesis entitled '' The secondary neutron sources for generation of particular neutron fluxes ''. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Swierk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Krakow (Poland). The physical problems to be solved in the two mentioned cases are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235 U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally. (author)

  15. Neutron generator power supply modeling in EMMA

    International Nuclear Information System (INIS)

    Robinson, A.C.; Farnsworth, A.V.; Montgomery, S.T.; Peery, J.S.; Merewether, K.O.

    1996-01-01

    Sandia National Laboratories has prime responsibility for neutron generator design and manufacturing, and is committed to developing predictive tools for modeling neutron generator performance. An important aspect of understanding component performance is explosively driven ferroelectric power supply modeling. EMMA (ElectroMechanical Modeling in ALEGRA) is a three dimensional compile time version of Sandia's ALEGRA code. The code is built on top of the general ALEGRA framework for parallel shock-physics computations but also includes additional capability for modeling the electric potential field in dielectrics. The overall package includes shock propagation due to explosive detonation, depoling of ferroelectric ceramics, electric field calculation and coupling with a general lumped element circuit equation system. The AZTEC parallel iterative solver is used to solve for the electric potential. The DASPK differential algebraic equation package is used to solve the circuit equation system. Sample calculations are described

  16. Assessing neutron generator output using neutron activation of silicon

    International Nuclear Information System (INIS)

    Kehayias, Pauli M.; Kehayias, Joseph J.

    2007-01-01

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the 28 Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10 3 n/s/cm 2 ± 5%, which is consistent with the manufacturer's specifications

  17. Assessing neutron generator output using neutron activation of silicon

    Energy Technology Data Exchange (ETDEWEB)

    Kehayias, Pauli M. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States); Kehayias, Joseph J. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States)]. E-mail: joseph.kehayias@tufts.edu

    2007-08-15

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the {sup 28}Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10{sup 3} n/s/cm{sup 2} {+-} 5%, which is consistent with the manufacturer's specifications.

  18. NEUTRON AND PHOTON DOSE MAPPING OF A DD NEUTRON GENERATOR.

    Science.gov (United States)

    Metwally, Walid A; Taqatqa, Osama A; Ballaith, Mohammed M; Chen, Allan X; Piestrup, Melvin A

    2017-11-01

    Neutron generators are an excellent tool that can be effectively utilized in educational institutions for applications such as neutron activation analysis, neutron radiography, and profiling and irradiation effects. For safety purposes, it is imperative that appropriate measures are taken in order to minimize the radiation dose from such devices to the operators, students and the public. This work presents the simulation and measurement results for the neutron and photon dose rates in the vicinity of the neutron generator installed at the University of Sharjah. A very good agreement is found between the simulated and measured dose rates. All of the public dose constraints were found to be met. The occupational dose constraint was also met after imposing a 200 cm no entry zone around the generator room. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  19. Compact neutron generator development and applications

    International Nuclear Information System (INIS)

    Leung, Ka-Ngo; Reijonen, Jani; Gicquel, Frederic; Hahto, Sami; Lou, Tak-Pui

    2004-01-01

    The Plasma and Ion Source Technology Group at the Lawrence Berkeley National Laboratory has been engaging in the development of high yield compact neutron generators for the last ten years. Because neutrons in these generators are formed by using either D-D, T-T or D-T fusion reaction, one can produce either mono-energetic (2.4 MeV or 14 MeV) or white neutrons. All the neutron generators being developed by our group utilize 13.5 MHz RF induction discharge to produce a pure deuterium or a mixture of deuterium-tritium plasma. As a result, ion beams with high current density and almost pure atomic ions can be extracted from the plasma source. The ion beams are accelerated to ∼100 keV and neutrons are produced when the beams impinge on a titanium target. Neutron generators with different configurations and sizes have been designed and tested at LBNL. Their applications include neutron activation analysis, oil-well logging, boron neutron capture therapy, brachytherapy, cargo and luggage screening. A novel small point neutron source has recently been developed for radiography application. The source size can be 2 mm or less, making it possible to examine objects with sharper images. The performance of these neutron generators will be described in this paper

  20. Measurement for skyshine of neutron generated by the K-600 neutron generator

    International Nuclear Information System (INIS)

    Zhen Huazhi; Li Guisheng; Wu Jingmin; Li Jianping

    1988-01-01

    The attenuation low of neutron scattering in atmosphere that generated by K-600 neutron generator at IMP was measured in order to evaluate the effect of the neutron generator to surroundings. The attenuation lenth λ = 396m was obtained and this result is in aggreement with the measured data at some laboratories abroad

  1. Compact neutron generator with nanotube ion source

    Science.gov (United States)

    Chepurnov, A. S.; Ionidi, V. Y.; Ivashchuk, O. O.; Kirsanov, M. A.; Kitsyuk, E. P.; Klenin, A. A.; Kubankin, A. S.; Nazhmudinov, R. M.; Nikulin, I. S.; Oleinik, A. N.; Pavlov, A. A.; Shchagin, A. V.; Zhukova, P. N.

    2018-02-01

    In this letter, we report the observation of fast neutrons generated when a positive acceleration potential is applied to an array of orientated carbon nanotubes, which are used as an ion source. The neutrons with energy of 2.45 MeV are generated as a result of D-D fusion reaction. The dependencies of the neutron yield on the value of the applied potential and residual pressure of deuterium are measured. The proposed approach is planned to be used for the development of compact neutron generators.

  2. Energy response of neutron area monitor with silicon semiconductor detector

    International Nuclear Information System (INIS)

    Kitaguchi, Hiroshi; Izumi, Sigeru; Kobayashi, Kaoru; Kaihara, Akihisa; Nakamura, Takashi.

    1993-01-01

    A prototype neutron area monitor with a silicon semiconductor detector has been developed which has the energy response of 1 cm dose equivalent recommended by the ICRP-26. Boron and proton radiators are coated on the surface of the silicon semiconductor detector. The detector is set at the center of a cylindrical polyethylene moderator. This moderator is covered by a porous cadmium board which serves as the thermal neutron absorber. Neutrons are detected as α-particles generated by the nuclear reaction 10 B(n,α) 7 Li and as recoil protons generated by the interaction of fast neutrons with hydrogen. The neutron energy response of the monitor was measured using thermal neutrons and monoenergetic fast neutrons generated by an accelerator. The response was consistent with the 1 cm dose equivalent response required for the monitor within ±34% in the range of 0.025 - 15 Mev. (author)

  3. Development of a Portable Fusion Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Byung-Hoon; In, Sang-Ryul; Jin, Jeong-Tae; Chang, Dae-Sik; Jang, Doh-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang Univ., Seoul (Korea, Republic of)

    2015-05-15

    For this purpose commercial ones, fast neutron yield from 10{sup 7} to 10{sup 11}, are supplied by several companies and research groups around the world. But internally the following limits make it difficult to develop the related application systems by domestic companies and/or research groups. - Limited life time - High price - Frequent trouble Not only to remove these limits but also to find out new internal application fields, it is necessary to develop our own domestic neutron generators. With the related technologies earned during fusion related researches, we did start to develop movable neutron generators from small one to big one, which could cover different fusion neutron yields. In this presentation the design and initial experimental results on the developed small neutron generator with a final target of 10{sup 8} n/s of 14 MeV neutrons, will be summarized.

  4. New portable neutron generator for well logging

    International Nuclear Information System (INIS)

    Chicanov, A.E.; Gromov, E. V.; Gulko, V. M.; Izmailov, A. V.

    1994-01-01

    The information about the design, investigation and testing of new well neutron generator for the pulse neutron logging (PNL) is given in this paper. The main physical characteristics of new PNL apparatus are: Neutron flux 2.10 sup 8 n/s ; Pulse frequency>=400 Hz; Diameter= 90 mm; Logging velocity >200 m/h; Number of probes = 2; Resource > 300 h. The generator were provided by gas-filled neutron accelerative tube named NTF-2. The perspective of application and optimization shown PNL apparatus are considered. (author)

  5. Characteristics of a Portable Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Jeong-Tae; Oh, Byung-Hoon; Chang, Dae-Sik; In, Sang-Yeol; Huh, Sung-Ryul; Hong, Kwang-Pyo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Neutron generators can be excellent tools for materials analysis, explosive material detection, nuclear weapon detection, and high quality radiography. D + D : 3He + n (2.5 MeV) D + T : 4He + n (14 MeV) Recent commercial neutron generators, fast neutron yield from 10{sup 7} to 10{sup 11} n/s, are produced by several companies and research groups around the world. But limited life time, high price, and frequent troubles make it difficult to develop related application systems by domestic companies or research groups. To remove such problems, it is necessary to develop our own domestic neutron generators. In this presentation, the design and experimental results on the developed small neutron generator are summarized. Experiments on deuterium beam extraction and fast neutron measurement by injecting deuterium beams on a drive-in target are executed. The stable deuterium beam of the energy higher than 100 keV was achieved by introducing metal cover which reduces the effect of metal-vacuum-insulator triple junction. The neutron flux of 5 n/s is measured by RadEye GN gamma Neutron (Thermo scientific) detector with about 200 mm distance and insertion of 40 mm PE plate between neutron source and the detector. The precise detector calibration is not carried out yet, so more detailed experimental results will be summarized at the presentation.

  6. Characteristics of a Portable Neutron Generator

    International Nuclear Information System (INIS)

    Jin, Jeong-Tae; Oh, Byung-Hoon; Chang, Dae-Sik; In, Sang-Yeol; Huh, Sung-Ryul; Hong, Kwang-Pyo

    2015-01-01

    Neutron generators can be excellent tools for materials analysis, explosive material detection, nuclear weapon detection, and high quality radiography. D + D : 3He + n (2.5 MeV) D + T : 4He + n (14 MeV) Recent commercial neutron generators, fast neutron yield from 10 7 to 10 11 n/s, are produced by several companies and research groups around the world. But limited life time, high price, and frequent troubles make it difficult to develop related application systems by domestic companies or research groups. To remove such problems, it is necessary to develop our own domestic neutron generators. In this presentation, the design and experimental results on the developed small neutron generator are summarized. Experiments on deuterium beam extraction and fast neutron measurement by injecting deuterium beams on a drive-in target are executed. The stable deuterium beam of the energy higher than 100 keV was achieved by introducing metal cover which reduces the effect of metal-vacuum-insulator triple junction. The neutron flux of 5 n/s is measured by RadEye GN gamma Neutron (Thermo scientific) detector with about 200 mm distance and insertion of 40 mm PE plate between neutron source and the detector. The precise detector calibration is not carried out yet, so more detailed experimental results will be summarized at the presentation

  7. Neutron Generators for Spent Fuel Assay

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard A.

    2010-01-01

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

  8. Neutronics activities for next generation devices

    International Nuclear Information System (INIS)

    Gohar, Y.

    1985-01-01

    Neutronic activities for the next generation devices are the subject of this paper. The main activities include TFCX and FPD blanket/shield studies, neutronic aspects of ETR/INTOR critical issues, and neutronics computational modules for the tokamak system code and tandem mirror reactor system code. Trade-off analyses, optimization studies, design problem investigations and computational models development for reactor parametric studies carried out for these activities are summarized

  9. New generation non-stationary portable neutron generators for biophysical applications of Neutron Activation Analysis.

    Science.gov (United States)

    Marchese, N; Cannuli, A; Caccamo, M T; Pace, C

    2017-01-01

    Neutron sources are increasingly employed in a wide range of research fields. For some specific purposes an alternative to existing large-scale neutron scattering facilities, can be offered by the new generation of portable neutron devices. This review reports an overview for such recently available neutron generators mainly addressed to biophysics applications with specific reference to portable non-stationary neutron generators applied in Neutron Activation Analysis (NAA). The review reports a description of a typical portable neutron generator set-up addressed to biophysics applications. New generation portable neutron devices, for some specific applications, can constitute an alternative to existing large-scale neutron scattering facilities. Deuterium-Deuterium pulsed neutron sources able to generate 2.5MeV neutrons, with a neutron yield of 1.0×10 6 n/s, a pulse rate of 250Hz to 20kHz and a duty factor varying from 5% to 100%, when combined with solid-state photon detectors, show that this kind of compact devices allow rapid and user-friendly elemental analysis. "This article is part of a Special Issue entitled "Science for Life" Guest Editor: Dr. Austen Angell, Dr. Salvatore Magazù and Dr. Federica Migliardo". Copyright © 2016 Elsevier B.V. All rights reserved.

  10. Neutron flux stabilization in the NG-150 neutron generators

    International Nuclear Information System (INIS)

    Kuz'min, L.E.; Makarov, S.A.; Pronman, I.M.

    1986-01-01

    Problem of metal tritium target lifetime increase and neutron flux stabilization in the NG-150 neutron generators is studied. Possibility on neutron flux stabilization using the mass analyzer for low-angle (4 deg and 41 deg) mass separation of a beam in thre components, which fall on a target simultaneously, is confirmed experimentally. Basic generator parameters are: accelerating voltage of 150 kV, total beam current on a target of 1.5 mA, beam current density of 0.3-1.6 mA/cm 2 , beam diameter of 8 mm. The initial neutron flux on the targets of 0.73 mg/cm 2 thick constituted 1.1x10 11 ssup(-1). The neutron flux monitoring was accomplished from recoil proton recording by a plastic scintillator. Flux decrease by more than 5% served as a signel for measuring mass analyzer magnetic field providing beam displacement on a target and restoration of the given flux. The NG-150 generator neutron flux stabilization was attained during 2h

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

    Science.gov (United States)

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

    2009-06-01

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

  12. 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)

  13. 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)

  14. Phototransistor response under a neutron fluence

    International Nuclear Information System (INIS)

    Santos, Luiz A.P.; Barros, Fabio R.; Ursulino, Luciano C.; Silva Junior, Eronides F.; Antonio Filho, Joao

    2009-01-01

    The purpose of this communication is to show some effects on a bipolar phototransistor after it has been under a neutron fluence. Unlike a transistor, a phototransistor is designed so that the collector has a large area and consequently it has a higher radiation detection probability. Then, it is possible to have a certain number of interactions so that any changes in the internal structure of the phototransistor can be observed after a neutron irradiation. If a phototransistor is under a certain spectra of neutron fluence the interaction depends on the cross section of the either silicon chip or its encapsulation, and recoil protons could be the charged particle responsible for changes in the semiconductor structure. Furthermore, neutron irradiation could give to the device a state of vanishing in its electrical characteristic which can be performed tracing the current versus voltage curve (I x V). The experimental arrangement basically consists of a photonic device, a neutron-gamma radiation source and a Flip-Flop electrometer second generation (EFF-2G). One of the main parameters of evaluation was the phototransistor dark current. In fact, the first results demonstrate that when the phototransistor is neutron irradiated there is a significant variation in its I x V characteristic curve. (author)

  15. 10 CFR 39.55 - Tritium neutron generator target sources.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

  16. Response of six neutron survey meters in mixed fields of fast and thermal neutrons.

    Science.gov (United States)

    Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S

    2013-10-01

    Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

  17. Neutron generator instrumentation at the Department 2350 Neutron Generator Test Facility

    International Nuclear Information System (INIS)

    Bryant, T.C.; Mowrer, G.R.

    1979-06-01

    The computer and waveform digitizing capability at the test facility has allowed several changes in the techniques used to test neutron generators. These changes include methods used to calibrate the instrumentation and changes in the operation of the test facility. These changes have increased the efficiency of the test facility as well as increasing both timing and amplitude accuracy of neutron generator waveforms

  18. Neutronic analysis of JET external neutron monitor response

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom (United Kingdom); Batistoni, Paola [ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden)

    2016-11-01

    Highlights: • We model JET tokamak containing JET remote handling system. • We investigate effect of remote handling system on external neutron monitor response. • Remote handling system correction factors are calculated. • Integral correction factors are relatively small, i.e up to 8%. - Abstract: The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated {sup 252}Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel. Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.

  19. Production of 14 MeV neutrons from D-D neutron generators

    International Nuclear Information System (INIS)

    Cecil, F.E.; Nieschmidt, E.B.

    1986-01-01

    The production of 14 MeV neutrons from a D-D neutron generator resulting from tritium buildup from the d(d,p)t reaction in the target is discussed. The effect of the 14 MeV neutrons on fast neutron activation analysis with D-D neutron generators is evaluated. (orig.)

  20. Fast neutron activation analysis by means of low voltage neutron generator

    Directory of Open Access Journals (Sweden)

    M.E. Medhat

    Full Text Available A description of D-T neutron generator (NG is presented. This machine can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. Procedure of neutron flux determination and efficiency calculation is described. Examples of testing some Egyptian natural cosmetics are given. Keywords: Neutron generator, Fast neutron activation analysis, Elemental analysis

  1. 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)

  2. Digitizing and analysis of neutron generator waveforms

    International Nuclear Information System (INIS)

    Bryant, T.C.

    1977-11-01

    All neutron generator waveforms from units tested at the SLA neutron generator test site are digitized and the digitized data stored in the CDC 6600 tape library for display and analysis using the CDC 6600 computer. The digitizing equipment consists mainly of seven Biomation Model 8100 transient recorders, Digital Equipment Corporation PDP 11/20 computer, RK05 disk, seven-track magnetic tape transport, and appropriate DEC and SLA controllers and interfaces. The PDP 11/20 computer is programmed in BASIC with assembly language drivers. In addition to digitizing waveforms, this equipment is used for other functions such as the automated testing of multiple-operation electronic neutron generators. Although other types of analysis have been done, the largest use of the digitized data has been for various types of graphical displays using the CDC 6600 and either the SD4020 or DX4460 plotters

  3. Plasma driven neutron/gamma generator

    Science.gov (United States)

    Leung, Ka-Ngo; Antolak, Arlyn

    2015-03-03

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

  4. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1980-01-01

    A specification is given for an energizing and regulating circuit for a gas filled neutron generator tube consisting of a target, an ion source and a replenisher, the circuit consisting of a power supply to provide a negative high voltage to the target and a target current corresponding to the neutron output of the tube, a constant current source, and control means connected to the power supply and to the constant current source, the control means being responsive to the target current to provide a portion of the constant current to the replenisher substantially to regulate the neutron output of the tube. (author)

  5. 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)

  6. 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.

  7. Neutron generator based on adiabatic trap

    International Nuclear Information System (INIS)

    Golovin, I.N.; Zhil'tsov, V.A.; Panov, D.A.; Skovoroda, A.A.; Shatalov, G.E.; Shcherbakov, A.G.

    1988-01-01

    A possibility of 14 MeV neutron generator (NG) production on the basis of axial-symmetric adiabatic trap with MHD cusped armature for the testing of materials and elements of the DT reactor first wall and blanket structure is discussed. General requirements to NG are formulated. It is shown that the NG variant discussed meets the requirements formulated. Approximate calculation of the NG parameters has shown that total energy consumption by the generator does not exceed 220 MW at neutron flux specific capacity of 2.5 MW/m 2 and radiation test area of 5-6 m 2

  8. Sustaining knowledge in the neutron generator community and benchmarking study.

    Energy Technology Data Exchange (ETDEWEB)

    Barrentine, Tameka C.; Kennedy, Bryan C.; Saba, Anthony W.; Turgeon, Jennifer L.; Schneider, Julia Teresa; Stubblefield, William Anthony; Baldonado, Esther

    2008-03-01

    In 2004, the Responsive Neutron Generator Product Deployment department embarked upon a partnership with the Systems Engineering and Analysis knowledge management (KM) team to develop knowledge management systems for the neutron generator (NG) community. This partnership continues today. The most recent challenge was to improve the current KM system (KMS) development approach by identifying a process that will allow staff members to capture knowledge as they learn it. This 'as-you-go' approach will lead to a sustainable KM process for the NG community. This paper presents a historical overview of NG KMSs, as well as research conducted to move toward sustainable KM.

  9. Secondary electron ion source neutron generator

    Science.gov (United States)

    Brainard, John P.; McCollister, Daryl R.

    1998-01-01

    A neutron generator employing an electron emitter, an ion source bombarded by the electrons from the electron emitter, a plasma containment zone, and a target situated between the plasma containment zone and the electron emitter. The target contains occluded deuterium, tritium, or a mixture thereof

  10. Utilization of a pulsed D-T neutron generator

    International Nuclear Information System (INIS)

    Vilaithong, T.; Singkarat, S.; Tippawan, U.

    2000-01-01

    In the past two decades the IAEA has supported the establishment of neutron laboratories in many developing countries by providing small D-T neutron generators. The neutron generator is basically a low energy (100-400 keV) ion accelerator capable of producing a continuous beam of deuterons with a current in the range between 1-2.5 mA. These neutron generators are primarily intended to be used for fast neutron activation analysis. This paper describes the utilization of a 14 MeV neutron generator in continuous and pulsed beam modes in applied neutron physics program at Chiang Mai University. (author)

  11. Characterization of Deuteron-Deuteron Neutron Generators

    Science.gov (United States)

    Waltz, Cory Scott

    A facility based on a next-generation, high-flux D-D neutron generator (HFNG) was commissioned at the University of California Berkeley. The characterization of the HFNG is presented in the following study. The current generator design produces near mono-energetic 2.45 MeV neutrons at outputs of 108 n/s. Calculations provided show that future conditioning at higher currents and voltages will allow for a production rate over 1010 n/s. Characteristics that effect the operational stability include the suppression of the target-emitted back streaming electrons, target sputtering and cooling, and ion beam optics. Suppression of secondary electrons resulting from the deuterium beam striking the target was achieved via the implementation of an electrostatic shroud with a voltage offset of greater than -400 V relative to the target. Ion beam optics analysis resulted in the creation of a defocussing extraction nozzle, allowing for cooler target temperatures and a more compact design. To calculate the target temperatures, a finite difference method (FDM) solver incorporating the additional heat removal effects of subcooled boiling was developed. Validation of the energy balance results from the finite difference method calculations showed the iterative solver converged to heat removal results within about 3% of the expected value. Testing of the extraction nozzle at 1.43 mA and 100 kV determined that overheating of the target did not occur as the measured neutron flux of the generator was near predicted values. Many factors, including the target stopping power, deuterium atomic species, and target loading ratio, affect the flux distribution of the HFNG neutron generator. A detailed analysis to understand these factors effects is presented. Comparison of the calculated flux of the neutron generator using deuteron depth implantation data, neutron flux distribution data, and deuterium atomic species data matched the experimentally calculated flux determined from indium foil

  12. High Intensity, Pulsed, D-D Neutron Generator

    International Nuclear Information System (INIS)

    Williams, D.L.; Vainionpaa, J.H.; Jones, G.; Piestrup, M.A.; Gary, C.K.; Harris, J.L.; Fuller, M.J.; Cremer, J.T.; Ludewigt, Bernhard A.; Kwan, J.W.; Reijonen, J.; Leung, K.-N.; Gough, R.A.

    2008-01-01

    Single ion-beam RF-plasma neutron generators are presented as a laboratory source of intense neutrons. The continuous and pulsed operations of such a neutron generator using the deuterium-deuterium fusion reaction are reported. The neutron beam can be pulsed by switching the RF plasma and/or a gate electrode. These generators are actively vacuum pumped so that a continuous supply of deuterium gas is present for the production of ions and neutrons. This contributes to the generator's long life. These single-beam generators are capable of producing up to 1E10 n/s. Previously, Adelphi and LBNL have demonstrated these generators applications in fast neutron radiography, Prompt Gamma Neutron Activation Analysis (PGNAA) and Neutron Activation Analysis (NAA). Together with an inexpensive compact moderator, these high-output neutron generators extend useful applications to home laboratory operations.

  13. 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

  14. Survey of Neutron Generators for Active Interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Moss, Calvin Elroy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sundby, Gary M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chichester, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Johnson, James P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-02

    Some of these commercially available generators meet all of the requirements in Table 1, but there are other concerns. Most generators containing SF6 will be required to have the SF6 gas removed for shipping because of DOT regulations. However, Thermo Fisher has a DOT exemption. The P211 and B211 from Thermo Fisher meet the requirements listed in Table 1, but they are old designs and are no longer offered for sale. Also, they require 15 minutes or more of warmup before neutron output is available, and they lack a modern digital control. The nGen-300C from Starfire Industries is interesting because it is a portable system, but it uses the DD reaction for 2.5 MeV neutrons, which are not as penetrating as the 14 MeV neutrons from the DT reaction. The MP 320 from Thermo Fisher is another portable system, but the minimum pulse rate is 250 Hz, which is too fast for measurement of delayed neutrons and re-interrogation by delayed neutrons between pulses. The Genie 16 from Sodern (from France) probably meets the requirements, but the required power is probably too high for battery operation. The generators from Russia and China may be difficult to purchase, and service may not be available. The power required by some of these generators is low enough that batteries can be used. The portable units, nGen-300C and the MP320, could easily be operated with batteries. Other generators with low power requirements, as specified in the above vendors list, could possibly be operated with reason size batteries. The batteries do not need to be internal to the generator, but can be in a separate package. The availability of high capacity lithium batteries with sophisticated safety circuits makes battery operation more possible now than when lead acid batteries were used. The best path forward probably requires working with vendors of the existing systems. If Starfire Industries could be persuaded to put tritium in their nGen-300C generator, possibly in collaboration with a national

  15. Development of a D-D Neutron Generator

    International Nuclear Information System (INIS)

    Kim, In Jung; Jung, Hwa Dong; Park, Chang Su; Jung, Nam Suk; Jung, Soon Wook; Hwang, Y. S.; Choi, H. D.

    2007-01-01

    To enhance neutron yield, the ion source of the D-D neutron generator is replaced by a large current helicon plasma ion source. Current and energy of deuteron beam are increased, and hence neutron yield is enhanced. The maximum neutron yield is 2x10 8 n/s

  16. Proposed Brookhaven accelerator-based neutron generator

    International Nuclear Information System (INIS)

    Grand, P.; Batchelor, K.; Chasman, R.; Rheaume, R.

    1976-01-01

    The d-Li Neutron Source concept, which includes a high-current dueteron linac, is an outgrowth of attempts made to use the BNL, 200-MeV proton linac BLIP facility to do radiation damage studies. It included a 100 mA, 30-MeV deuteron linear accelerator and a fast-flowing liquid lithium jet as the target. The latest design is not very different, except that the current is now 200 mA and the linac energy has been raised to 35 MeV. Both parameters, were changed to optimize the effectiveness of the facility with respect to flux, experimental volume and match to 14 MeV neutron-radiation-damage effects. The proposed Brookhaven Accelerator-based Neutron Generator is described with particular emphasis on the linear accelerator. The proposed facility is a practical and efficient way of producing the intense, high energy neutron beams needed for CTR material studies. The accelerator and liquid-metal technologies are well proven, state-of-the-art technologies. The fact that no new technology is required guarantees the possibility of meeting construction schedules, and more importantly, guarantees a high level of operational reliability

  17. Neutron PSDs for the next generation of spallation neutron sources

    CERN Document Server

    Eijk, C W

    2002-01-01

    A review of R and D for neutron PSDs to be used at anticipated new spallation neutron sources: the Time-of-Flight system facility, European Spallation Source, Spallation Neutron Source and Neutron Arena, is presented. The gas-filled detectors, scintillation detectors and hybrid systems are emphasized.

  18. SIMULATED 8 MeV NEUTRON RESPONSE FUNCTIONS OF A THIN SILICON NEUTRON SENSOR.

    Science.gov (United States)

    Takada, Masashi; Matsumoto, Tetsuro; Masuda, Akihiko; Nunomiya, Tomoya; Aoyama, Kei; Nakamura, Takashi

    2017-12-22

    Neutron response functions of a thin silicon neutron sensor are simulated using PHITS2 and MCNP6 codes for an 8 MeV neutron beam at angles of incidence of 0°, 30° and 60°. The contributions of alpha particles created from the 28Si(n,α)25Mg reaction and the silicon nuclei scattered elastically by neutrons in the silicon sensor have not been well reproduced using the MCNP6 code. The 8 MeV neutron response functions simulated using the PHITS2 code with an accurate event generator mode are in good agreement with experimental results and include the contributions of the alpha particles and silicon nuclei. © The Author(s) 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  19. Response of electret dosemeter to slow neutrons

    International Nuclear Information System (INIS)

    Ghilardi, A.J.P.; Pela, C.A.; Zimmerman, R.L.

    1987-01-01

    The response of electret dosemeter to slow neutrons exposure is cited, mentioning the preparation and the irradiation of dosemeter with Am-Be source. Some theory considerations about the response of electret dosemeter to slow and fast neutrons are also presented. (C.G.C.) [pt

  20. D-D neutron generator development at LBNL.

    Science.gov (United States)

    Reijonen, J; Gicquel, F; Hahto, S K; King, M; Lou, T-P; Leung, K-N

    2005-01-01

    The plasma and ion source technology group in Lawrence Berkeley National Laboratory is developing advanced, next generation D-D neutron generators. There are three distinctive developments, which are discussed in this presentation, namely, multi-stage, accelerator-based axial neutron generator, high-output co-axial neutron generator and point source neutron generator. These generators employ RF-induction discharge to produce deuterium ions. The distinctive feature of RF-discharge is its capability to generate high atomic hydrogen species, high current densities and stable and long-life operation. The axial neutron generator is designed for applications that require fast pulsing together with medium to high D-D neutron output. The co-axial neutron generator is aimed for high neutron output with cw or pulsed operation, using either the D-D or D-T fusion reaction. The point source neutron generator is a new concept, utilizing a toroidal-shaped plasma generator. The beam is extracted from multiple apertures and focus to the target tube, which is located at the middle of the generator. This will generate a point source of D-D, T-T or D-T neutrons with high output flux. The latest development together with measured data will be discussed in this article.

  1. Pathways to agility in the production of neutron generators

    Energy Technology Data Exchange (ETDEWEB)

    Stoltz, R.E. [Sandia National Labs., Livermore, CA (United States); Beavis, L.C.; Cutchen, J.T.; Garcia, P.; Gurule, G.A.; Harris, R.N.; McKey, P.C.; Williams, D.W. [Sandia National Labs., Albuquerque, NM (United States)

    1994-02-01

    This report is the result of a study team commissioned to explore pathways for increased agility in the manufacture of neutron generators. As a part of Sandia`s new responsibility for generator production, the goal of the study was to identify opportunities to reduce costs and increase flexibility in the manufacturing operation. Four parallel approaches (or pathways) were recommended: (1) Know the goal, (2) Use design leverage effectively, (3) Value simplicity, and (4) Configure for flexibility. Agility in neutron generator production can be enhanced if all of these pathways are followed. The key role of the workforce in achieving agility was also noted, with emphasis on ownership, continuous learning, and a supportive environment.

  2. Overview of the Division 2351 Neutron Generator Test Facility waveform digitizing system. [Explosively activated neutron generators

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, T.C. Jr.

    1978-02-01

    All neutron generator waveforms from units tested at the SLA neutron generator test site are digitized and the digitized data stored in the CDC 6600 tape library for display and analysis using the CDC 6600 computer. The digitizing equipment consists mainly of seven Biomation Model 8100 transient recorders, Digital Equipment Corporation PDP 11/20 computer, RK05 disk, seven-track magnetic tape transport, and appropriate DEC and SLA controllers and interfaces. The PDP 11/20 computer is programmed in BASIC with assembly language drivers. In addition to digitizing waveforms, this equipment is used for other functions such as the automated testing of multiple-operation electronic neutron generators. Although other types of analysis have been done, the largest use of the digitized data has been for various types of graphical displays using the CDC 6600 and either the SD4020 or DX4460 plotters.

  3. A search for solar neutron response in neutron monitor data

    International Nuclear Information System (INIS)

    Kudela, K.

    1990-01-01

    The search for an impulsive increase corresponding to a solar neutron response on high-mountain neutron monitors requires control of the stability of the measurement and elimination of other sources of short-time increases of different kinds which are involved in fluctuations of cosmic-ray intensity. For the solar flare of June 3, 1982 the excess of counting rate on the Lomnicky stit neutron monitor is, within a factor or 1.8, equal to that expected from solar neutrons. Superposed epoch analysis of 17 flares with gamma-ray or hard X-ray production gives a slight tendency of an occurring signal in cases of high heliocentric angles, indicating anisotropic production of neutrons on the sun. The low statistical significance of the result indicates that higher temporal resolution, better evaluation of multiplicity, better knowledge of the power spectra of short-term intensity fluctuations on neutron monitors, as well as coordinated measurements of solar gamma-rays and neutrons on satellites, are required. 21 refs

  4. Study of general digital DC/pulse neutron generator

    International Nuclear Information System (INIS)

    Li Gang; Liu Zheng; Li Wensheng; Liu Hanlin; Liu Linmao

    2014-01-01

    Preliminary experimental results of digital DC/pulse neutron generator based on a ceramic drive-in target neutron tube for explosives detection are presented. The generator is a portable and on-off neutron source, and it can be controlled by remote PC. The generator can produce DC neutrons, pulse neutrons and multiple pulse neutrons. The maximum neutron yield is about 2 × 10"8 n/s, the minimum pulse width is 10 μs and the maximum pulse frequency is 10 kHz. Neutron yield and time-spectrum is measured in China Academy of Engineering Physics. The generator is suitable for explosive detection with PFTNA technology, and it can be used in other areas such as reactor measurements and on-line industrial test systems. (authors)

  5. Method and apparatus for generating neutrons

    International Nuclear Information System (INIS)

    Cranberg, L.

    1978-01-01

    An apparatus and method for generating high-energy neutrons are disclosed. Neutron emissive target material is deposited on one or more surfaces on a rotatable, hollow, toroidal target support. The surfaces are bombarded by beams of ions of generally rectangular cross section, so that when the bombarded surfaces are viewed end-wise, a compact, generally square source of neutrons is provided, such as is required for collimation. A combination of molecular and atomic ions emitted from at least one conventional accelerator are passed through a magnetic field for the purpose of separating the ions into one homogeneous group of atomic and one homogeneous group of molecular ions before the ions are allowed to impinge on the target surfaces. One accelerator directs ions to each target surface as the target rotates. Coolant is directed through a cavity within the toroidal support for the purpose of cooling the target support and target material. A refrigerated surface is placed in close proximity to the target surface to condense vapors which might prove harmful to the target and for thermally cooling said target

  6. Neutron generator tube ion source control apparatus

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

    A pulsed neutron well logging system includes a neutron generator tube of the deuterium-tritium accelerator type and an ion source control apparatus providing extremely sharply time-defined neutron pulses. A low voltage control pulse supplied to an input by timing circuits turns a power FET on via a buffer-driver whereby a 2000 volt pulse is produced in the secondary of a pulse transformer and applied to the ion source of the tube. A rapid fall in this ion source control pulse is ensured by a quenching circuit wherein a one-shot responds to the falling edge of the control pulse and produces a 3 microsecond delay to compensate for the propagation delay. A second one-shot is triggered by the falling edge of the output of the first one-shot and gives an 8 microsecond pulse to turn on the power FET which, via an isolation transformer turns on a series-connected transistor to ground the secondary of the pulse transformer and the ion source. (author)

  7. Study on the energy response to neutrons for a new scintillating-fiber-array neutron detector

    CERN Document Server

    Zhang Qi; Wang Qun; Xie Zhong Shen

    2003-01-01

    The energy response of a new scintillating-fiber-array neutron detector to neutrons in the energy range 0.01 MeV<=E sub n<=14 MeV was modeled by combining a simplified Monte Carlo model and the MCNP 4b code. In order to test the model and get the absolute sensitivity of the detector to neutrons, one experiment was carried out for 2.5 and 14 MeV neutrons from T(p,n) sup 3 He and T(d,n) sup 4 He reactions at the Neutron Generator Laboratory at the Institute of Modern Physics, the Chinese Academy of Science. The absolute neutron fluence was obtained with a relative standard uncertainty 4.5% or 2.0% by monitoring the associated protons or sup 4 He particles, respectively. Another experiment was carried out for 0.5, 1.0, 1.5, 2.0, 2.5 MeV neutrons from T(p,n) sup 3 He reaction, and for 3.28, 3.50, 4.83, 5.74 MeV neutrons from D(d,n) sup 3 He reaction on the Model 5SDH-2 accelerator at China Institute of Atomic Energy. The absolute neutron fluence was obtained with a relative standard uncertainty 5.0% by usin...

  8. 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

  9. Neutron energy response measurement of scintillation detectors

    International Nuclear Information System (INIS)

    Yang Hongqiong; Peng Taiping; Yang Jianlun; Tang Zhengyuan; Yang Gaozhao; Li Linbo; Hu Mengchun; Wang Zhentong; Zhang Jianhua; Li Zhongbao; Wang Lizong

    2004-01-01

    Neutron sensitivities of detectors composed of plastic scintillator ST401, ST1422, ST1423 and phyotomultiplier tube in primary energy range of fission neutron are calibrated by direct current. The energy response curve of the detectors is obtained in this experiment. The experimental result has been compared with the theoretical calculation and they are in agreement within measuring uncertainty. (authors)

  10. Development of fast neutron radiography system based on portable neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Chia Jia, E-mail: gei-i-kani@hotmail.com; Nilsuwankosit, Sunchai, E-mail: sunchai.n@chula.ac.th [Department of Nuclear Engineering, Faculty of Engineering, Chulalongkorn University, Phayathai Rd., Patumwan, Bangkok, THAILAND 10330 (Thailand)

    2016-01-22

    Due to the high installation cost, the safety concern and the immobility of the research reactors, the neutron radiography system based on portable neutron generator is proposed. Since the neutrons generated from a portable neutron generator are mostly the fast neutrons, the system is emphasized on using the fast neutrons for the purpose of conducting the radiography. In order to suppress the influence of X-ray produced by the neutron generator, a combination of a shielding material sandwiched between two identical imaging plates is used. A binary XOR operation is then applied for combining the information from the imaging plates. The raw images obtained confirm that the X-ray really has a large effect and that XOR operation can help enhance the effect of the neutrons.

  11. Indoor Fast Neutron Generator for Biophysical and Electronic Applications

    Science.gov (United States)

    Cannuli, A.; Caccamo, M. T.; Marchese, N.; Tomarchio, E. A.; Pace, C.; Magazù, S.

    2018-05-01

    This study focuses the attention on an indoor fast neutron generator for biophysical and electronic applications. More specifically, the findings obtained by several simulations with the MCNP Monte Carlo code, necessary for the realization of a shield for indoor measurements, are presented. Furthermore, an evaluation of the neutron spectrum modification caused by the shielding is reported. Fast neutron generators are a valid and interesting available source of neutrons, increasingly employed in a wide range of research fields, such as science and engineering. The employed portable pulsed neutron source is a MP320 Thermo Scientific neutron generator, able to generate 2.5 MeV neutrons with a neutron yield of 2.0 x 106 n/s, a pulse rate of 250 Hz to 20 KHz and a duty factor varying from 5% to 100%. The neutron generator, based on Deuterium-Deuterium nuclear fusion reactions, is employed in conjunction with a solid-state photon detector, made of n-type high-purity germanium (PINS-GMX by ORTEC) and it is mainly addressed to biophysical and electronic studies. The present study showed a proposal for the realization of a shield necessary for indoor applications for MP320 neutron generator, with a particular analysis of the transport of neutrons simulated with Monte Carlo code and described the two main lines of research in which the source will be used.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  13. Development of compact D-D neutron generator

    International Nuclear Information System (INIS)

    Das, Basanta Kumar; Das, Rashmita; Shyam, Anurag

    2011-12-01

    In recent years, due to specific features of compact neutron generators, their demand in elemental analysis and detection of the illicit materials has been increased in scientific community. Compact is size, controlled operation and radiation safety like features of neutron generator is suitable for research work with illicit materials. An accelerator based neutron generator can be operated in steady mode as well as in pulse mode. The main embodiment of this type of generator includes ion source, ion acceleration system and target. We are developing such type of neutron generator. This consists of one-in-house developed penning ion source, a single electrode acceleration gap and one deuterated titanium target or virgin titanium target. In this report, we will discuss various physics and technical issues related to the important components of this generator, operation of the generator and neutron detection. (author)

  14. Generation of laser-induced fast neutron and its application

    International Nuclear Information System (INIS)

    Cha, Hyung Ki; Lee, S.; Kwon, D.; Nam, S.; Park, S.; Rhee, Y.; Jung, Y.; Lee, K.; Cha, Y.; Kwon, S.; Lim, C.; Han, J.; Park, S.; Chung, C.

    2012-04-01

    The supply of high-efficiency neutron source is still problematic even though a fast neutron source is being accepted increasingly for industrial applications. Radioisotopes and a neutron tube are typically being used, but their neutron flux, lifetime, and price are the limiting factors for more diverse applications. As ultra high power, short pulse laser technologies have been developed, a neutron source generated via laser induced nuclear reaction comes to the fore. The laser induced neutron source has a high peak flux in comparison to the traditional neutron source and is like a point source with its diameter less than 1 mm. These properties can be utilized effectively for the analysis of pulsed fast neutron activation or the studies of a fast neutron material damage and/or recover. The purpose of R and D here is to develop a robust neutron source with a yield of 107 neutrons/s during 1st R and D stage ('07 ∼ '09) and to construct a stable laser neutron source in longer operation and to demonstrate its usefulness for a neutron activation analysis of explosive materials and a neutron impact analysis of crystalline in the second R and D stage ('10 ∼ '11)

  15. Fusion neutron detector calibration using a table-top laser generated plasma neutron source

    International Nuclear Information System (INIS)

    Hartke, R.; Symes, D.R.; Buersgens, F.; Ruggles, L.E.; Porter, J.L.; Ditmire, T.

    2005-01-01

    Using a high intensity, femtosecond laser driven neutron source, a high-sensitivity neutron detector was calibrated. This detector is designed for observing fusion neutrons at the Z accelerator in Sandia National Laboratories. Nuclear fusion from laser driven deuterium cluster explosions was used to generate a clean source of nearly monoenergetic 2.45 MeV neutrons at a well-defined time. This source can run at 10 Hz and was used to build up a clean pulse-height spectrum on scintillating neutron detectors giving a very accurate calibration for neutron yields at 2.45 MeV

  16. Neutron multiplication in lead in the experiments with neutron generators

    International Nuclear Information System (INIS)

    Markovskij, D.V.

    1989-01-01

    A calculational analysis of neutron multiplication in lead, including the estimates of multiplication limits for the standard ENDF/BIV data set and the effects of various changes in the data themselves is performed. 10 refs, 5 figs

  17. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Science.gov (United States)

    Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  18. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, Sean [Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Ghani, Zamir [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, Igor [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Milocco, Alberto; Packer, Lee [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, Mario [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  19. Development of a sealed-accelerator-tube neutron generator

    Science.gov (United States)

    Verbeke; Leung; Vujic

    2000-10-01

    Sealed-accelerator-tube neutron generators are being developed in Lawrence Berkeley National Laboratory (LBNL) for applications ranging from neutron radiography to boron neutron capture therapy and neutron activation analysis. The new generation of high-output neutron generators is based on the D-T fusion reaction, producing 14.1-MeV neutrons. The main components of the neutron tube--the ion source, the accelerator and the target--are all housed in a sealed metal container without external pumping. Thick-target neutron yield computations are performed in this paper to estimate the neutron yield of titanium and scandium targets. With an average deuteron beam current of 1 A and an energy of 120 keV, a time-averaged neutron production of approximately 10(14) n/s can be estimated for a tritiated target, for both pulsed and cw operations. In mixed deuteron/triton beam operation, a beam current of 2 A at 150 keV is required for the same neutron output. Recent experimental results on ion sources and accelerator columns are presented and discussed.

  20. Analysis of the neutron generation from a D-Li neutron source

    International Nuclear Information System (INIS)

    Gomes, I.

    1994-02-01

    The study of the neutron generation from the D-Li reaction is an important issue to define the optimum combination of the intervening parameters during the design phase of a D-Li neutron source irradiation facility. The major players in defining the neutron yield from the D-Li reaction are the deuteron incident energy and the beam current, provided that the lithium target is thick enough to stop all incident deuterons. The incident deuteron energy also plays a role on the angular distribution of the generated neutrons, on the energy distribution of the generated neutrons, and on the maximum possible energy of the neutrons. The D-Li reaction produces neutrons with energies ranging from eV's to several MeV's. The angular distribution of these neutrons is dependent on the energy of both, incident deuterons and generated neutrons. The deuterons lose energy interacting with the lithium target material in such a way that the energy of the deuterons inside the lithium target varies from the incident deuteron energy to essentially zero. The first part of this study focuses in analyzing the neutron generation rate from the D-Li reaction as a function of the intervening parameters, in defining the source term, in terms of the energy and angular distributions of the generated neutrons, and finally in providing some insights of the impact of varying input parameters on the generation rate and correlated distributions. In the second part an analytical description of the Monte Carlo sampling procedure of the neutron from the D-Li reaction is provided with the aim at further Monte Carlo transport of the D-Li neutrons

  1. 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

  2. The intensive DT neutron generator of TU Dresden

    Science.gov (United States)

    Klix, Axel; DÖring, Toralf; Domula, Alexander; Zuber, Kai

    2018-01-01

    TU Dresden operates an accelerator-based intensive DT neutron generator. Experimental activities comprise investigation into material activation and decay, neutron and photon transport in matter and R&D work on radiation detectors for harsh environments. The intense DT neutron generator is capable to produce a maximum of 1012 n/s. The neutron source is a solid-type water-cooled tritium target based on a titanium matrix on a copper carrier. The neutron yield at a typical deuteron beam current of 1 mA is of the order of 1011 n/s in 4Π. A pneumatic sample transport system is available for short-time irradiations and connected to wo high-purity germanium detector spectrometers for the measurement of induced activities. The overall design of the experimental hall with the neutron generator allows a flexible setup of experiments including the possibility of investigating larger structures and cooled samples or samples at high temperatures.

  3. 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.

  4. Mechanism of neutron generation in Z-pinches

    International Nuclear Information System (INIS)

    Vikhrev, V.V.

    1986-01-01

    The review of experimental and theoretical investigations in a mechanism of neutron generation in Z-pinches is presented. Special attention is paid to the thermonuclear mechanism of neutron generation occuring due to the formation of high-temperature plasma regions in Z-pinch sausage-type instabilities. This mechanism is shown to be predominant in charges with the neutron yield more than 10 9 per a charge. Experimental data, which are considered to be contradicting to thermonuclear nature of neutron radiation, are explained

  5. Radiological safety aspects of the operation of neutron generators

    International Nuclear Information System (INIS)

    Boggs, R.F.

    1976-01-01

    The purpose of the manual is to provide some basic guidelines to persons with a minimum of training in radiological health or health physics, on some safety aspects of the operation of sealed-tube and Cockcroft-Walton type neutron generators. The manual does not state rules or regulations but presents a description of the most likely hazards. It is relevant to those relatively compact neutron generators which usually operate at less than 150-200 kV for the purpose of producing 14-MeV neutrons. The scope is limited to basic discussions of hazards and measurement techniques. Separate chapters are devoted to the characteristics and use of neutron generators; radiation hazards and safety considerations; radiation monitoring and interpretation of measurements; and requirements for an effective safety programme. Two appendices deal with non-radiation hazards and safety considerations, and with a neutron generator laboratory, respectively. An extensive list of bibliographic references is included

  6. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1980-01-01

    A gas filled neutron tube in a nuclear well logging tool has a target an ion source voltage and a replenisher connected to ground. A negative high voltage is applied to the target by a power supply also providing a target current corresponding to the neutron output of the neutron generator tube. A constant current source provides a constant current. A network receiving the target current and the constant current provides a portion of the constant current as a replenisher current which is applied to the replenisher in a neutron generating tube. The network controls the magnitude of the replenisher current in accordance with the target current so as to control the neutron output of the neutron generating tube. (auth)

  7. Production of radionuclides by 14 MeV neutron generator

    International Nuclear Information System (INIS)

    Alfassi, Z.B.

    1983-01-01

    Due to the short half-lives of these nuclides they have to be produced in situ or at least not far from the place of use. The cost of 14 MeV neutron generators have been compared with the typical middle-sized cyclotrons and it was found that the capital costs are much lower in the case of neutron generators. This is the main reason for the availability of 14 MeV neutron generators in many scientific institutes compared to the scarcity of cyclotrons. Lately, the use of 14 MeV neutrons for cancer therapy was studied in several medical centers. A number of hospitals and cancer research centers have high intensity 14 MeV neutron generators for this purpose. The advantages of using short-lived in-house produced radionuclides suggest the use of the available 14 MeV neutron generators for biological studies and in medical diagnosis. 14 MeV neutron generators can be used to produce some of the medically useful radionuclides, such as /sup 18/F, /sup 80/Br, /sup 199m/Hg, and others. However, the amount required for medicine can only be prepared by the new high intensity neutron generators, used for neutron therapy and not by the smaller ones, commonly used in university laboratories (--10/sup 11/ n/sec). On the other hand, these relatively small neutron generators can be used for the preparation of radionuclides for biological studies. They facilitate the study of metabolism of elements for which radionuclides cannot be usually purchased due to short half-lives or the high price of the long-lived ones, such as /sup 34m/Cl, /sup 18/F, /sup 28,29/Al, /sup 27/Mg, and others. An example is the work done on the fate of Al and Mg in rats using /sup 28/Al and /sup 27/Mg./sup 13/

  8. Fission multipliers for D-D/D-T neutron generators

    International Nuclear Information System (INIS)

    Lou, T.P.; Vujic, J.L.; Koivunoro, H.; Reijonen, J.; Leung, K.-N.

    2003-01-01

    A compact D-D/D-T fusion based neutron generator is being designed at the Lawrence Berkeley National Laboratory to have a potential yield of 10 12 D-D n/s and 10 14 D-T n/s. Because of its high neutron yield and compact size (∼20 cm in diameter by 4 cm long), this neutron generator design will be suitable for many applications. However, some applications required higher flux available from nuclear reactors and spallation neutron sources operated with GeV proton beams. In this study, a subcritical fission multiplier with k eff of 0.98 is coupled with the compact neutron generators in order to increase the neutron flux output. We have chosen two applications to show the gain in flux due to the use of fission multipliers--in-core irradiation and out-of-core irradiation. For the in-core irradiation, we have shown that a gain of ∼25 can be achieved in a positron production system using D-T generator. For the out-of-core irradiation, a gain of ∼17 times is obtained in Boron Neutron Capture Therapy (BNCT) using a D-D neutron generator. The total number of fission neutrons generated by a source neutron in a fission multiplier with k eff is ∼50. For the out-of-core irradiation, the theoretical maximum net multiplication is ∼30 due to the absorption of neutrons in the fuel. A discussion of the achievable multiplication and the theoretical multiplication will be presented in this paper

  9. Generation of laser-induced fast neutron and its application

    International Nuclear Information System (INIS)

    Cha, Hyung Ki; Kwon, D. H.; Nam, S. M.

    2010-04-01

    The supply of high-efficiency neutron source is still problematic even though a fast neutron source is being accepted increasingly for industrial applications. Radioisotopes and a neutron tube are typically being used, but their neutron flux, lifetime, and price are the limiting factors for more diverse applications. As ultra high power, short pulse laser technologies have been developed, a neutron source generated via laser induced nuclear reaction comes to the fore. The laser induced neutron source has a high peak flux in comparison to the traditional neutron source and is like a point source with its diameter less than 1 mm. These properties can be utilized effectively for the analysis of pulsed fast neutron activation or the studies of a fast neutron material damage and/or recover. The purpose of R and D here is to develop a robust neutron source with a yield of 10 7 neutrons/s, and to carry out a preliminary research for application study in the next research stage

  10. Response functions of superfluid neutron matter

    Energy Technology Data Exchange (ETDEWEB)

    Keller, Jochen; Sedrakian, Armen [Institut fuer Theoretische Physik, Goethe Universitaet, 60438 Frankfurt/Main (Germany)

    2011-07-01

    We investigate the response of pair-correlated neutron matter under conditions relevant to neutron stars to external weak probes and compute its neutrino emissivity in vector and axialvector channels. To derive the response functions we sum up an infinite chain of particle-hole ladder diagrams within finite-temperature Green's function theory. The polarization tensor of matter is evaluated in the limit of small momentum transfers. The calculated neutrino emission via the weak neutral current processes of pair-breaking and recombination of Cooper-pairs in neutron stars causes a cooling of their baryonic interior, and represents an important mechanism for the thermal evolution of the star within a certain time domain.

  11. Research and development activities of a neutron generator facility

    International Nuclear Information System (INIS)

    Darsono Sudjatmoko; Pramudita Anggraita; Sukarman Aminjoyo

    2000-01-01

    The neutron generator facility at YNRC is used for elemental analysis, nuclear data measurement and education. In nuclear data measurement the focus is on re-evaluating the existing scattered nuclear activation cross-section to obtain systematic data for nuclear reactions such as (n,p), (n,α), and (n,2n). In elemental analysis it is used for analyzing the Nitrogen (N), Phosphor (P) and Potassium (K) contents in chemical and natural fertilizers (compost), protein in rice, soybean, and corn and pollution level in rivers. The neutron generator is also used for education and training of BATAN staff and university students. The facility can also produce neutron generator components. (author)

  12. Response characteristics of selected personnel neutron dosimeters

    International Nuclear Information System (INIS)

    McDonald, J.C.; Fix, J.J.; Hadley, R.T.; Holbrook, K.L.; Yoder, R.C.; Roberson, P.L.; Endres, G.W.R.; Nichols, L.L.; Schwartz, R.B.

    1983-09-01

    Performance characteristics of selected personnel neutron dosimeters in current use at Department of Energy (DOE) facilities were determined from their evaluation of neutron dose equivalent received after irradiations with specific neutron sources at either the National Bureau of Standards (NBS) or the Pacific Northwest Laboratory (PNL). The characteristics assessed included: lower detection level, energy response, precision and accuracy. It was found that when all of the laboratories employed a common set of calibrations, the overall accuracy was approximately +-20%, which is within uncertainty expected for these dosimeters. For doses above 80 mrem, the accuracy improved to better than 10% when a common calibration was used. Individual differences found in this study may reflect differences in calibration technique rather than differences in the dose rates of actual calibration standards. Second, at dose rates above 100 mrem, the precision for the best participants was generally below +-10% which is also within expected limits for these types of dosimeters. The poorest results had a standard deviation of about +-25%. At the lowest doses, which were sometimes below the lower detection limit, the precision often approached or exceeded +-100%. Third, the lower level of detection for free field 252 Cf neutrons generally ranged between 20 and 50 mrem. Fourth, the energy dependence study provided a characterization of the response of the dosimeters to neutron energies far from the calibration energy. 11 references, 22 figures, 26 tables

  13. Generating energy dependent neutron flux maps for effective ...

    African Journals Online (AJOL)

    For activation analysis and irradiation scheme of miniature neutron source reactor, designers or engineers usually require information on thermal neutron flux levels and other energy group flux levels (such as fast, resonance and epithermal). A methodology for readily generating such flux maps and flux profiles for any ...

  14. Research possibilities with an intense neutron generator

    International Nuclear Information System (INIS)

    Bartholomew, G.A.

    1966-01-01

    As the title suggests this paper will depart somewhat from the general topic of this session and will be concerned more with applications of accelerators than with accelerators them elves. The particular application of interest at our laboratory concerns the use of a high current intermediate energy proton accelerator as the basis for a versatile intense neutron source. Chalk River's entry into the intermediate energy accelerator field with neutron production as the primary motivation is somewhat unusual. Although neutron production is also being explored by other laboratories interested in intermediate energy accelerators, e.g., Oak Ridge National Laboratory and Los Alamos Scientific Laboratory, it has not been the major motivation. Our initial motivation was in fact the production of thermal neutrons and this interest has remained foremost in our ING program. We are currently writing a proposal for this project. Our target is to have a proton beam in 1973. (author)

  15. Research possibilities with an intense neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Bartholomew, G A

    1966-07-01

    As the title suggests this paper will depart somewhat from the general topic of this session and will be concerned more with applications of accelerators than with accelerators them elves. The particular application of interest at our laboratory concerns the use of a high current intermediate energy proton accelerator as the basis for a versatile intense neutron source. Chalk River's entry into the intermediate energy accelerator field with neutron production as the primary motivation is somewhat unusual. Although neutron production is also being explored by other laboratories interested in intermediate energy accelerators, e.g., Oak Ridge National Laboratory and Los Alamos Scientific Laboratory, it has not been the major motivation. Our initial motivation was in fact the production of thermal neutrons and this interest has remained foremost in our ING program. We are currently writing a proposal for this project. Our target is to have a proton beam in 1973. (author)

  16. Note: Coincidence measurements of 3He and neutrons from a compact D-D neutron generator

    Science.gov (United States)

    Ji, Q.; Lin, C.-J.; Tindall, C.; Garcia-Sciveres, M.; Schenkel, T.; Ludewigt, B. A.

    2017-05-01

    Tagging of neutrons (2.45 MeV) with their associated 3He particles from deuterium-deuterium (D-D) fusion reactions has been demonstrated in a compact neutron generator setup enabled by a high brightness, microwave-driven ion source with a high fraction of deuterons. Energy spectra with well separated peaks of the D-D fusion reaction products, 3He, tritons, and protons, were measured with a silicon PIN diode. The neutrons were detected using a liquid scintillator detector with pulse shape discrimination. By correlating the 3He detection events with the neutron detection in time, we demonstrated the tagging of emitted neutrons with 3He particles detected with a Si PIN diode detector mounted inside the neutron generator vacuum vessel.

  17. Response of electret dosemeter to slow neutrons

    International Nuclear Information System (INIS)

    Ghilardi, A.J.P.; Pela, C.A.; Zimmerman, R.L.

    1987-01-01

    The response of the electret dosemeter to exposition of slow neutrons is studied. Different external coatings are used on the dosemeter (polyethylene, alminium, polyethylene + boron, aluminium + boron) and exposure curves (with and without water) are compared. (M.A.C.) [pt

  18. Development of high flux thermal neutron generator for neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

    2015-05-01

    The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

  19. The neutron production rate measurement of an indigenously developed compact D-D neutron generator

    Directory of Open Access Journals (Sweden)

    Das Basanta Kumar

    2013-01-01

    Full Text Available One electrostatic accelerator based compact neutron generator was developed. The deuterium ions generated by the ion source were accelerated by one accelerating gap after the extraction from the ion source and bombarded to a target. Two different types of targets, the drive - in titanium target and the deuteriated titanium target were used. The neutron generator was operated at the ion source discharge potential at +Ve 1 kV that generates the deuterium ion current of 200 mA at the target while accelerated through a negative potential of 80 kV in the vacuum at 1.3×10-2 Pa filled with deuterium gas. A comparative study for the neutron yield with both the targets was carried out. The neutron flux measurement was done by the bubble detectors purchased from Bubble Technology Industries. The number of bubbles formed in the detector is the direct measurement of the total energy deposited in the detector. By counting the number of bubbles the total dose was estimated. With the help of the ICRP-74 neutron flux to dose equivalent rate conversion factors and the solid angle covered by the detector, the total neutron flux was calculated. In this presentation the operation of the generator, neutron detection by bubble detector and estimation of neutron flux has been discussed.

  20. Feasibility of sealed D-T neutron generator as neutron source for liver BNCT and its beam shaping assembly.

    Science.gov (United States)

    Liu, Zheng; Li, Gang; Liu, Linmao

    2014-04-01

    This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. Copyright © 2014 Elsevier Ltd. All rights reserved.

  1. Portable Neutron Generator with 9-Section Silicon $\\alpha $-Detector

    CERN Document Server

    Bystritsky, V M; Kadyshevskij, V G; Khasaev, T O; Kobzev, A P; Presnyakov, Yu K; Rogov,Yu N; Ryzhkov, V I; Sapozhnikov, M G; Sissakian, A N; Slepnev, V M; Zamyatin, N I

    2006-01-01

    The characteristics of the portable neutron generator with a built-in $\\alpha $-detector are presented. Based on the "tagged" neutron method (TNM) the generator ~is being used for identification of ~the hidden chemical compounds. One of the special features of such generators compared to generators traditionally used and produced in industry is that the generator is a source of monoenergetic "tagged" 14.1 MeV neutrons produced in the binary nuclear reaction $d+t \\to \\alpha $ (3.5 MeV) $+n$ (14.1~MeV). Unambiguous information about the time and direction of the neutron emitted from the target can be obtained by recording an $\\alpha $ particle by the multi-pixel $\\alpha $-detector placed inside the neutron tube. The study of the "tagged" neutron method (TNM) shows that the use of the ($\\alpha $--$\\gamma $) coincidence reduces the gamma background induced by scattered neutrons by a factor of more than 200, which allows the detection and identification of small quantities of explosives, drugs, and toxic agents. T...

  2. High yield neutron generators using the DD reaction

    Energy Technology Data Exchange (ETDEWEB)

    Vainionpaa, J. H.; Harris, J. L.; Piestrup, M. A.; Gary, C. K.; Williams, D. L.; Apodaca, M. D.; Cremer, J. T. [Adelphi technology, 2003 E. Bayshore Rd. 94061, Redwood City, CA (United States); Ji, Qing; Ludewigt, B. A. [Lawrence Berkeley National Lab, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Jones, G. [G and J Enterprise, 1258 Quary Ln, Suite F, Pleasanton California 94566 (United States)

    2013-04-19

    A product line of high yield neutron generators has been developed at Adelphi technology inc. The generators use the D-D fusion reaction and are driven by an ion beam supplied by a microwave ion source. Yields of up to 5 Multiplication-Sign 10{sup 9} n/s have been achieved, which are comparable to those obtained using the more efficient D-T reaction. The microwave-driven plasma uses the electron cyclotron resonance (ECR) to produce a high plasma density for high current and high atomic ion species. These generators have an actively pumped vacuum system that allows operation at reduced pressure in the target chamber, increasing the overall system reliability. Since no radioactive tritium is used, the generators can be easily serviced, and components can be easily replaced, providing essentially an unlimited lifetime. Fast neutron source size can be adjusted by selecting the aperture and target geometries according to customer specifications. Pulsed and continuous operation has been demonstrated. Minimum pulse lengths of 50 {mu}s have been achieved. Since the generators are easily serviceable, they offer a long lifetime neutron generator for laboratories and commercial systems requiring continuous operation. Several of the generators have been enclosed in radiation shielding/moderator structures designed for customer specifications. These generators have been proven to be useful for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA) and fast neutron radiography. Thus these generators make excellent fast, epithermal and thermal neutron sources for laboratories and industrial applications that require neutrons with safe operation, small footprint, low cost and small regulatory burden.

  3. Development and characterization of a D-D fast neutron generator for imaging applications.

    Science.gov (United States)

    Adams, Robert; Bort, Lorenz; Zboray, Robert; Prasser, Horst-Michael

    2015-02-01

    The experimental characterization of a pulsed D-D fast neutron generator designed for fan-beam tomography applications is presented. Using Monte Carlo simulations the response of an LB6411 neutron probe was related to the neutron generator output. The yield was measured to be up to ∼10(7) neutrons/s. An aluminum block was moved stepwise between the source and a BC400 plastic scintillator detector in order to measure an edge response. This edge response was related to the neutron emitting spot size using Monte Carlo simulations and a simplified geometry-based model. The experimentally determined spot size of 2.2 mm agreed well with the simulated value of 1.5 mm. The time-dependence of pulsed output for various operating conditions was also measured. The neutron generator was found to satisfy design requirements for a planned fast neutron tomography arrangement based on a plastic scintillator detector array which is expected to be capable of producing 2D tomograms with a resolution of ∼1.5 mm. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. 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)

  5. A single-beam deuteron compact accelerator for neutron generation

    International Nuclear Information System (INIS)

    Araujo, Wagner Leite; Campos, Tarcisio Passos Ribeiro de

    2011-01-01

    Portable neutron generators are devices composed by small size accelerators that produce neutrons through fusion between hydrogen isotopes. These reactions are characterized by appreciable cross section at energies at the tens of keV, which enables device portability. The project baselines follow the same physical and engineering principles of any other particle accelerators. The generator consists of a gas reservoir, apparatus for ion production, few electrodes to accelerate and focus the ion beam, and a metal hydride target where fusion reactions occur. Neutron generator applications include geophysical measurements, indus- trial process control, environmental, research, nation's security and mechanical structure analysis.This article presents a design of a compact accelerator for d-d neutron generators, describing the physical theory applied to the deuteron extraction system, and simulating the ion beam transport in the accelerator. (author)

  6. Elemental analysis using temporal gating of a pulsed neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Mitra, Sudeep

    2018-02-20

    Technologies related to determining elemental composition of a sample that comprises fissile material are described herein. In a general embodiment, a pulsed neutron generator periodically emits bursts of neutrons, and is synchronized with an analyzer circuit. The bursts of neutrons are used to interrogate the sample, and the sample outputs gamma rays based upon the neutrons impacting the sample. A detector outputs pulses based upon the gamma rays impinging upon the material of the detector, and the analyzer circuit assigns the pulses to temporally-based bins based upon the analyzer circuit being synchronized with the pulsed neutron generator. A computing device outputs data that is indicative of elemental composition of the sample based upon the binned pulses.

  7. Development of a compact D-D neutron generator

    Science.gov (United States)

    Huang, Z.-W.; Wang, J.-R.; Wei, Z.; Lu, X.-L.; Ma, Z.-W.; Ran, J.-L.; Zhang, Z.-M.; Yao, Z.-E.; Zhang, Y.

    2018-01-01

    A compact D-D neutron generator was developed at Lanzhou University, China. A duoplasmatron ion source was used to produce a higher-current deuteron beam. The deuteron beam could be accelerated up to 150 keV by a single accelerating gap, and bombarded on a pure molybdenum drive-in target to produce D-D fast neutron. A bias voltage between the target and the extraction-accelerating electrode was produced by a resistance to suppress the secondary electron from the target. The neutron generator has been operated for several hundred hours, and the performances were investigated. The available range of the deuteron beam current was 1.0-4.0 mA. EJ410 scintillator detector system was used to measure the fast neutron yields. D-D neutron yield could reach 2.48×108 n/s under the deuteron beam of 3 mA and 150 keV.

  8. A neutron dynamic therapy with a boron tracedrug UTX-51 using a compact neutron generator.

    Science.gov (United States)

    Hori, Hitoshi; Tada, Ryu; Uto, Yoshihiro; Nakata, Eiji; Morii, Takashi; Masuda, Kai

    2014-08-01

    We are developing a neutron dynamic therapy (NDT) with boron tracedrugs for a new mechanical-clearance treatment of pathotoxic misfolded, aggregated, and self-propagating prion-associated disease proteins. We present a compact neutron generator-based NDT using a boron tracedrug UTX-51. Our NDT is based on the weak thermal neutron-bombarded destructive action of UTX-51 on bovine serum albumin (BSA) using the neutron beams produced from a compact inertial electrostatic confinement fusion (IECF) neutron generator. BSA as an NDT molecular target was subjected to thermal neutron irradiation for eight hours using a compact neutron generator. The sodium dodecyl sulfate-polyacrylamide gel electrophoresis pattern showed no protein band when 2 nmoles of BSA were irradiated with more than 100 nmoles of UTX-51, while BSA was not affected when irradiated without UTX-51. For the first time, we have succeeded in the molecular destruction of a prion-disease model protein, BSA, by NDT with a boron tracedrug, UTX-51, using a compact neutron generator. Copyright© 2014 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

  9. Target injection and engagement for neutron generation at 1 Hz

    International Nuclear Information System (INIS)

    Komeda, Osamu; Mori, Yoshitaka; Nishimura, Yasuhiko

    2013-01-01

    Target injection is a key technology to realizing inertial fusion energy. Here we present the first demonstration of target injection and neutron generation. We injected more than 600 spherical deuterated polystyrene (C 8 D 8 ) bead targets during 10 minutes at 1 Hz. After the targets fell for a distance of 18 cm, we applied the synchronized laser-diode-pumped ultra-intense laser HAMA and successfully generated neutrons repeatedly. The result is a step toward fusion power and also suggests possible industrial neutron sources. (author)

  10. 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

  11. Pulsed White Spectrum Neutron Generator for Explosive Detection

    International Nuclear Information System (INIS)

    King, Michael J.; Miller, Gill T.; Reijonen, Jani; Ji, Qing; Andresen, Nord; Gicquel, Frederic; Kavlas, Taneli; Leung, Ka-Ngo; Kwan, Joe

    2008-01-01

    Successful explosive material detection in luggage and similar sized containers is a critical issue in securing the safety of all airline passengers. Tensor Technology Inc. has recently developed a methodology that will detect explosive compounds with pulsed fast neutron transmission spectroscopy. In this scheme, tritium beams will be used to generate neutrons with a broad energy spectrum as governed by the T(t,2n)4He fission reaction that produces 0-9 MeV neutrons. Lawrence Berkeley National Laboratory (LBNL), in collaboration with Tensor Technology Inc., has designed and fabricated a pulsed white-spectrum neutron source for this application. The specifications of the neutron source are demanding and stringent due to the requirements of high yield and fast pulsing neutron emission, and sealed tube, tritium operation. In a unique co-axial geometry, the ion source uses ten parallel rf induction antennas to externally couple power into a toroidal discharge chamber. There are 20 ion beam extraction slits and 3 concentric electrode rings to shape and accelerate the ion beam into a titanium cone target. Fast neutron pulses are created by using a set of parallel-plate deflectors switching between +-1500 volts and deflecting the ion beams across a narrow slit. The generator is expected to achieve 5 ns neutron pulses at tritium ion beam energies between 80-120 kV. First experiments demonstrated ion source operation and successful beam pulsing

  12. Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

    International Nuclear Information System (INIS)

    Saito, Masaki; Artisyuk, V.; Chmelev, A.

    1999-01-01

    The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

  13. Investigation of the response characteristics of OSL albedo neutron dosimeters in a 241AmBe reference neutron field

    Science.gov (United States)

    Liamsuwan, T.; Wonglee, S.; Channuie, J.; Esoa, J.; Monthonwattana, S.

    2017-06-01

    The objective of this work was to systematically investigate the response characteristics of optically stimulated luminescence Albedo neutron (OSLN) dosimeters to ensure reliable personal dosimetry service provided by Thailand Institute of Nuclear Technology (TINT). Several batches of InLight® OSLN dosimeters were irradiated in a reference neutron field generated by the in-house 241AmBe neutron irradiator. The OSL signals were typically measured 24 hours after irradiation using the InLight® Auto 200 Reader. Based on known values of delivered neutron dose equivalent, the reading correction factor to be used by the reader was evaluated. Subsequently, batch homogeneity, dose linearity, lower limit of detection and fading of the OSLN dosimeters were examined. Batch homogeneity was evaluated to be 0.12 ± 0.05. The neutron dose response exhibited a linear relationship (R2=0.9974) within the detectable neutron dose equivalent range under test (0.4-3 mSv). For this neutron field, the lower limit of detection was between 0.2 and 0.4 mSv. Over different post-irradiation storage times of up to 180 days, the readings fluctuated within ±5%. Personal dosimetry based on the investigated OSLN dosimeter is considered to be reliable under similar neutron exposure conditions, i.e. similar neutron energy spectra and dose equivalent values.

  14. Characterization of a deuterium-deuterium plasma fusion neutron generator

    Science.gov (United States)

    Lang, R. F.; Pienaar, J.; Hogenbirk, E.; Masson, D.; Nolte, R.; Zimbal, A.; Röttger, S.; Benabderrahmane, M. L.; Bruno, G.

    2018-01-01

    We characterize the neutron output of a deuterium-deuterium plasma fusion neutron generator, model 35-DD-W-S, manufactured by NSD/Gradel-Fusion. The measured energy spectrum is found to be dominated by neutron peaks at 2.2 MeV and 2.7 MeV. A detailed GEANT4 simulation accurately reproduces the measured energy spectrum and confirms our understanding of the fusion process in this generator. Additionally, a contribution of 14 . 1 MeV neutrons from deuterium-tritium fusion is found at a level of 3 . 5%, from tritium produced in previous deuterium-deuterium reactions. We have measured both the absolute neutron flux as well as its relative variation on the operational parameters of the generator. We find the flux to be proportional to voltage V 3 . 32 ± 0 . 14 and current I 0 . 97 ± 0 . 01. Further, we have measured the angular dependence of the neutron emission with respect to the polar angle. We conclude that it is well described by isotropic production of neutrons within the cathode field cage.

  15. Photon response of silicon diode neutron detectors

    International Nuclear Information System (INIS)

    McCall, R.C.; Jenkins, T.M.; Oliver, G.D. Jr.

    1976-07-01

    The photon response of silicon diode neutron detectors was studied to solve the problem on detecting neutrons in the presence of high energy photons at accelerator neutron sources. For the experiment Si diodes, Si discs, and moderated activation foil detectors were used. The moderated activation foil detector consisted of a commercial moderator and indium foils 2'' in diameter and approximately 2.7 grams each. The moderator is a cylinder of low-density polyethylene 6 1 / 4 '' in diameter by 6 1 / 16 '' long covered with 0.020'' of cadmium. Neutrons are detected by the reaction 115 In (n,γ) 116 In(T/sub 1 / 2 / = 54 min). Photons cannot be detected directly but photoneutrons produced in the moderator assembly can cause a photon response. The Si discs were thin slices of single-crystal Si about 1.4 mils thick and 1'' in diameter which were used as activation detectors, subsequently being counted on a thin-window pancake G.M. counter. The Si diode fast neutron dosimeter 5422, manufactured by AB Atomenergi in Studsvik, Sweden, consists of a superdoped silicon wafer with a base width of 0.050 inches between two silver contacts coated with 2 mm of epoxy. For this experiment, the technique of measuring the percent change of voltage versus dose was used. Good precision was obtained using both unirradiated and preirradiated diodes. All diodes, calibrated against 252 CF in air,were read out 48 hours after irradiation to account for any room temperature annealing. Results are presented and discussed

  16. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1978-01-01

    Means and method are described for energizing and regulating a neutron generator tube having a target, an ion source and a replenisher. It providing a negative high voltage to the target and monitoring the target current. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordence with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current

  17. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1977-01-01

    A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

  18. Spectral fluence of neutrons generated by radiotherapeutic Linacs

    International Nuclear Information System (INIS)

    Kralik, Miloslav; Solc, Jaroslav; Smoldasova, Jana; Vondracek, Vladimir; Farkasova, Estera; Ticha, Ivana

    2015-01-01

    Spectral fluences of neutrons generated in the heads of the radiotherapeutic linacs Varian Clinac 2100 C/D and Siemens ARTISTE were measured by means of the Bonner spheres spectrometer whose active detector of thermal neutrons was replaced by an activation detector, i.e. a tablet made of pure manganese. Measurements with different collimator settings reveal an interesting dependence of neutron fluence on the area defined by the collimator jaws. The determined neutron spectral fluences were used to derive ambient dose equivalent rate along the treatment coach. To clarify at which components of the linac neutrons are mainly created, the measurements were complemented with MCNPX calculations based on a realistic model of the Varian Clinac. (authors)

  19. Energy response study of modified CR-39 neutron personnel dosimeter

    International Nuclear Information System (INIS)

    Sathian, Deepa; Bakshi, A.K.; Datta, D.; Nair, Sreejith S.; Sathian, V.; Mishra, Jitendra; Sen, Meghnath

    2018-01-01

    Personnel neutron dosimetry is an integral part of radiation protection. No single dosimeter provides the satisfactory energy response, sensitivity, angular dependence characteristics and accuracy necessary to meet the requirement of an ideal personnel neutron dosimeter. The response of a personnel neutron dosimeter is critically dependent upon the energy distribution of the neutron field. CR-39 personnel neutron dosimeters were typically calibrated in the standard neutron field of 252 Cf and 241 Am-Be in our laboratory, although actual neutron fields may vary from the calibration neutron spectrum. Recently the badge cassette of the personnel neutron dosimeter was changed due to frequent damage of the PVC badge used earlier. This paper discusses energy response of CR-39 solid state nuclear track detector loaded in this modified badge cassette as per latest ISO recommendation

  20. Tritium contamination and monitoring at Frascati Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Lucci, F.; Sandri, S.; Ianni, A. [ENEA, Frascati (Italy). Dipartimento Ambiente; Vasselli, R. [ANPA, Roma (Italy); Pillon, M.; Bettinali, L. [ENEA, Frascati (Italy). Dipartimento Energia

    1994-11-01

    The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10{sup 1}1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed.

  1. Tritium contamination and monitoring at Frascati Neutron Generator

    International Nuclear Information System (INIS)

    Lucci, F.; Sandri, S.; Ianni, A.; Pillon, M.; Bettinali, L.

    1994-11-01

    The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10 1 1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed

  2. Modification of 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; Park, Ji Young; Hwang, Yong Seok; Choi, H. D.

    2005-01-01

    The prototype D-D neutron generator was modified in order to enhance the neutron yield. The distance from ion source to target was reduced to increase the ion beam current at target position. Thick Ti target was replaced by thin Ti target which was vacuum-deposited on Cu substrate in order to enhance the target cooling. Performance of the modified device was tested

  3. On generating neutron transport tables with the NJOY system

    International Nuclear Information System (INIS)

    Caldeira, Alexandre D.; Claro, Luiz H.

    2013-01-01

    Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)

  4. The Response of Alanine Dosimeters in Thermal Neutron Fields

    DEFF Research Database (Denmark)

    Schmitz, T.; Bassler, Niels; Sharpe, P.

    response of all pellets could be reproduced by calculations within a uncertainty of 5 %. For all experiments three dose components have been separated. A proton dose is generated in the 14N(n,p)14C reaction. Secondary gammas are generated by various (n,γ) reactions, dominated by the 2.2 MeV photon from...... experiments the dosimeters will be exposed to higher neutron energies, which are more typical for BNCT treatments. References: [1] Barth, R.F; 2009: Boron neutron capture therapy at the crossroads: Challenges and opportunities. Applied Radiation and Isotopes 67, 3-6. [2] Rogus, R.D.; Harling, O.K.; Yanch, J.C...... for treatment of liver metastases. Applied Radiation and Isotopes 67, 238-241. [4] Sharpe, P.; Sephtan, J.; 2000: An automated system for the measurement of alanine/EPR dosimeters. Applied Radiation and Isotopes 52, 1185-1188....

  5. A Monte-Carlo method for ex-core neutron response

    International Nuclear Information System (INIS)

    Gamino, R.G.; Ward, J.T.; Hughes, J.C.

    1997-10-01

    A Monte Carlo neutron transport kernel capability primarily for ex-core neutron response is described. The capability consists of the generation of a set of response kernels, which represent the neutron transport from the core to a specific ex-core volume. This is accomplished by tagging individual neutron histories from their initial source sites and tracking them throughout the problem geometry, tallying those that interact in the geometric regions of interest. These transport kernels can subsequently be combined with any number of core power distributions to determine detector response for a variety of reactor Thus, the transport kernels are analogous to an integrated adjoint response. Examples of pressure vessel response and ex-core neutron detector response are provided to illustrate the method

  6. The intensive DT neutron generator of TU Dresden

    Directory of Open Access Journals (Sweden)

    Klix Axel

    2018-01-01

    Full Text Available TU Dresden operates an accelerator-based intensive DT neutron generator. Experimental activities comprise investigation into material activation and decay, neutron and photon transport in matter and R&D work on radiation detectors for harsh environments. The intense DT neutron generator is capable to produce a maximum of 1012 n/s. The neutron source is a solid-type water-cooled tritium target based on a titanium matrix on a copper carrier. The neutron yield at a typical deuteron beam current of 1 mA is of the order of 1011 n/s in 4Π. A pneumatic sample transport system is available for short-time irradiations and connected to wo high-purity germanium detector spectrometers for the measurement of induced activities. The overall design of the experimental hall with the neutron generator allows a flexible setup of experiments including the possibility of investigating larger structures and cooled samples or samples at high temperatures.

  7. Online monitoring of fast neutron (DT/DD) at Purnima neutron generator

    International Nuclear Information System (INIS)

    Bishnoi, S.; Patel, T.; Shukla, M.; Adhikari, P.S.; Sinha, A.

    2012-01-01

    A neutron generator (NG) at Purnima Labs, BARC has been developed for DT accelerator driven zero power subcritical (ADSS) system. Subcritical core of ADSS will be coupled to the NG for benchmarking experiments. Kinetic parameters of ADSS such as K-source, flux, power etc depends on this external neutron source strength injected to the core. However the neutron emission rate of NG does not remain stable throughout its operation. In view of this a reliable, precise and online monitoring of NG's neutron emission rate is required. An online neutron monitoring system based on associated particle method has been designed, developed and installed at NG. The monitoring unit consists of an ion implanted planar silicon detector, placed inside the drift tube of NG at an angle with respect to D + beam direction. A series of experiments were carried out with increasing neutron yield to optimize the position of detector such that it has sufficient counting statistics and minimum pileup. A complementary calibration procedure for validating these results based on activation technique was also carried out with standard Cu foil. The reaction rate monitored with online monitor and foil activation technique were compared, their variations with the predicted (theoretical) results were within 16%. This paper deals with the development and performance of online neutron monitoring system for DT and DD neutrons

  8. Spatial distribution of neutron flux for the A-711 neutron generator

    International Nuclear Information System (INIS)

    Essiet, A. E.; Owolabi, S. A.; Adesanmi, C. A.; Balogun, F. A.

    1996-01-01

    The spatial distribution of neutron flux for the Kaman sciences A-711 neutron generator recently installed at the Centre for Energy Research and Development (CERD), Ile-Ife Nigeria has been determined. At an operational tube current of 2.0 mA and high voltage power supply (HVPS) of 158 kV, the neutron flux increases from 1.608 ± 0.021*10 8 n/cm 2 s at the top of the irradiated plastic vial to 2.640 ± 0.022*10 8 n/cm 2 s at the centre, and then decreases to 1.943 ± 0.02* 8 n/cm 2 s at the bottom. The flux density is strongly dependent on the diameter of deuteron at the tritium target, and within this range a source strength of 10 8 n/s has been measured for the A-711 neutron generator

  9. Calculation of Spectra of Neutrons and Charged Particles Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    An algorithm for calculating the spectra of neutrons and associated charged particles produced in the target of a neutron generator is detailed. The products of four nuclear reactions 3H( d, n)4He, 2H( d, n)3He, 2H( d, p)3H, and 3He( d, p)4He are analyzed. The results of calculations are presented in the form of neutron spectra for several emission angles and spectra of associated charged particles emitted at an angle of 180° for a deuteron initial energy of 0.13 MeV.

  10. Neutron lifetime and generation time by KENO IV

    International Nuclear Information System (INIS)

    Hayashi, Masatoshi

    1991-01-01

    It is believed that Monte Carlo method is suitable to the calculation of neutron lifetime and generation time with reference to the life cycle viewpoint. This paper illustrates that those times obtained by Monte Carlo method are quite different from the results by perturbation method. The neutron lifetime and the generation time for bare and reflected reactors were investigated by the Monte Carlo program, KENO IV. the Monte Carlo procedure is based on tracking and recording the life history of neutrons in a realistic fashion in a fissionable system with minimum nuclear and geometric approximations. The KENO IV provides the multiplication factor, neutron lifetime and generation time simultaneously. The thermal spherical reactors for both bare and reflected reactors were studied using the KENO IV. The reflected reactor is surrounded with 30 cm thick light water. The atomic densities in the regions and the calculated results of the multiplication factor, neutron lifetime and generation time are given. The different definitions of these times between the Monte Carlo method and perturbation theory caused the difference of the results. (K.I.)

  11. Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

    International Nuclear Information System (INIS)

    Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

    2011-01-01

    Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

  12. The shielding of a 14 MeV neutron generator

    International Nuclear Information System (INIS)

    Brighton, D.R.

    1976-10-01

    The concrete masonry shield for a 14 MeV neutron generator was designed using data supplied by the manufacturer. Subsequent radiation surveys outside the shield showed doses higher than expected. Calculations indicated the sensitivity of dose transmission factors to concrete composition. The observed dose transmission factor agreed with that of Broerse but not with that of Hacke and Prudhomme. Measurements and calculations delineated the contribution that neutrons, scattered from the upper wall that supports the laboratory roof, made to the dose in adjoining areas. In redesigning the shield a compromise was made between additional cost and restrictions on the generator's duty cycle, which is automatically controlled to ensure personnel safety. (Author)

  13. Beam splitting to improve target life in neutron generators

    International Nuclear Information System (INIS)

    Farrell, J.P.

    1976-01-01

    In a neutron generator in which a tritium-titanium target is bombarded by a deuterium ion beam, the target half-life is increased by separating the beam with a weak magnetic field to provide three separate beams of atomic, diatomic, and triatomic deuterium ions which all strike the target at different adjacent locations. Beam separation in this manner eliminates the problem of one type ion impairing the neutron generating efficiency of other type ions, thereby effecting more efficient utilization of the target material

  14. ITEP Subcritical Neutron Generator driven by charged particle accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Shvedov, O.V.; Chuvilo, I.V.; Vasiliev, V.V. [Institute of Theoretical and Experimental Physics, Moscow (Russian Federation)] [and others

    1995-10-01

    A research facility prototype including a combination of a linear accelerator, a neutron generating target, a nuclear safety ensuring and means of its attainment for Subcritical Neutron Generator are considered. The scheme of the multiplying is shown. The assembly will be mounted in the body of the partly dismantled ITEP HWR. Requirements for subcritical assembly are worked out and their feasibility within the framework of the heavy-water blanket is shown. The facility`s application as a full-scale model of more powerful installations of this kind and for fundamental experimental research has been investigated.

  15. 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.

  16. The intense neutron generator and future factory type ion accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1968-07-01

    A neutron factory is likely to sell its product in the form of isotopes. To ay neutron factories are nuclear reactors. Ion accelerators may also produce isotopes by direct interaction and, at high enough energies, mesons and hyperons. The challenge of the electrical production of neutrons goes far beyond the isotope market. It challenges the two popular concepts for long term large scale energy, the fast breeder reactor and controlled thermonuclear fusion. For this use about 4% of nuclear generated power would be applied in a feedback loop generating extra neutrons. Competition rests on operating and processing costs. The Intense Neutron Generator proposal now cancelled would have been full scale for such a use, but much further advance in accelerator engineering is required and anticipated. Perhaps most promising is the application of the ion drag principle in which rings of fast electrons are accelerated along their axis dragging ions with them by electrostatic attraction. Due to the much larger mass of the ions they can acquire much higher energy than the electrons and the process could be efficient. Such accelerators have not yet been made but experimental and theoretical studies are promising. (author)

  17. The intense neutron generator and future factory type ion accelerators

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1968-01-01

    A neutron factory is likely to sell its product in the form of isotopes. To ay neutron factories are nuclear reactors. Ion accelerators may also produce isotopes by direct interaction and, at high enough energies, mesons and hyperons. The challenge of the electrical production of neutrons goes far beyond the isotope market. It challenges the two popular concepts for long term large scale energy, the fast breeder reactor and controlled thermonuclear fusion. For this use about 4% of nuclear generated power would be applied in a feedback loop generating extra neutrons. Competition rests on operating and processing costs. The Intense Neutron Generator proposal now cancelled would have been full scale for such a use, but much further advance in accelerator engineering is required and anticipated. Perhaps most promising is the application of the ion drag principle in which rings of fast electrons are accelerated along their axis dragging ions with them by electrostatic attraction. Due to the much larger mass of the ions they can acquire much higher energy than the electrons and the process could be efficient. Such accelerators have not yet been made but experimental and theoretical studies are promising. (author)

  18. Multi detector input and function generator for polarized neutron experiments

    International Nuclear Information System (INIS)

    De Blois, J.; Beunes, A.J.H.; Ende, P. v.d.; Osterholt, E.A.; Rekveldt, M.T.; Schipper, M.N.; Velthuis, S.G.E. te

    1998-01-01

    In this paper a VME module is described for static or stroboscopic measurements with a neutron scattering instrument, consisting essentially of a series of up to 64 3 He neutron detectors around a sample environment. Each detector is provided with an amplifier and a discriminator to separate the neutrons from noise. To reduce the wiring, the discriminator outputs are connected to the module by coding boxes. Two 16-inputs to one-output coding boxes generate serial output codes on a fiber optic connection. This basically fast connection reduces the dead time introduced by the coding, and the influence of environmental noise. With stroboscopic measurements a periodic function is used to affect the sample surrounded by a field coil. Each detected neutron is labeled with a data label containing the detector number and the time of detection with respect to a time reference. The data time base can be programmed on a linear or a nonlinear scale. An external source or an attribute of the periodic function may generate the time reference pulse. A 12-bit DAC connected to the output of an 8 K, 16-bits memory, where the pattern of the current has been stored before, generates the function. The function memory is scanned by the programmable function time base. Attributes are set by the four remaining bits of the memory. One separate detector input connects a monitor detector in the neutron beam with a 32-bit counter/timer that provides measuring on a preset count, preset time or preset frame. (orig.)

  19. Compact Neutron Generators for Medical, Home Land Security, and Planetary Exploration

    CERN Document Server

    Reijonen, Jani

    2005-01-01

    The Plasma and Ion Source Technology Group at Lawrence Berkeley National Laboratory has developed various types of advanced D-D (neutron energy 2.5 MeV), D-T (14 MeV) and T-T (0 - 9 MeV) neutron generators for wide range of applications. These applications include medical (Boron Neutron Capture Therapy), homeland security (Prompt Gamma Activation Analysis, Fast Neutron Activation Analysis and Pulsed Fast Neutron Transmission Spectroscopy) and planetary exploration in form of neutron based, sub-surface hydrogen detection systems. These neutron generators utilize RF induction discharge to ionize the deuterium/tritium gas. This discharge method provides high plasma density for high output current, high atomic species from molecular gases, long life operation and versatility for various discharge chamber geometries. Three main neutron generator developments are discussed here: high neutron output co-axial neutron generator for BNCT applications, point neutron generator for security applications, compact and sub-c...

  20. Neutron radiation characteristics of the IVth generation reactor spent fuel

    Science.gov (United States)

    Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

    2018-03-01

    Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

  1. Radiation protection aspects of a high flux, fast neutron generator

    International Nuclear Information System (INIS)

    DeLuca, P.M.; Torti, R.P.; Chenevert, G.M.; Tesmer, J.R.; Kelsey, C.A.

    1976-01-01

    During the development and operation of a gas target, DT neutron generator for use in cancer therapy, two radiation hazards were routinely encountered - personnel exposure to neutrons and to tritium. The principal hazard was irradiation by fast neutrons. By assembling the source below ground level, adding shielding and the use of a controlled access, key identification interlock, the neutron hazard has been reduced. With the present source strength of 2 x 10 12 n/sec, an average neutron dose rate in the control room of 20 mrem/hr was measured- a level compatible with a limited run schedule. The second hazard was exposure to tritium in both gaseous and solid forms. A target inventory of 90 Ci, and overall inventory of 500 Ci, and the need to modify and repair the generator present significant potential hazard due to tritium exposure. The use of protective gloves, wipe tests, urine assays, continuous room air monitoring, and equipment decontamination minimized personnel exposure and effectively confined contamination. The dose due to tritium has been ∼ 0.5 rem/year and negligible spread of contamination has occurred

  2. Manual for troubleshooting and upgrading of neutron generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    This manual is intended to assist operators in troubleshooting and upgrading of neutron generators. It is directed particularly to operators and technicians in less experienced laboratories and therefore the descriptions of the principles and techniques of these machines are operator oriented. In addition to a discussion of the main characteristics of neutron generators, detailed information is given on the function of particular commercial units, on common problems related to specific components of accelerators, and on methods of troubleshooting and repair. Detailed schematic and circuit diagrams are provided to help operators in the development and improvement of the generators. The problems treated in the Manual have been collected during several IAEA missions in developing countries. 125 refs, 161 figs, 22 tabs.

  3. Manual for troubleshooting and upgrading of neutron generators

    International Nuclear Information System (INIS)

    1996-11-01

    This manual is intended to assist operators in troubleshooting and upgrading of neutron generators. It is directed particularly to operators and technicians in less experienced laboratories and therefore the descriptions of the principles and techniques of these machines are operator oriented. In addition to a discussion of the main characteristics of neutron generators, detailed information is given on the function of particular commercial units, on common problems related to specific components of accelerators, and on methods of troubleshooting and repair. Detailed schematic and circuit diagrams are provided to help operators in the development and improvement of the generators. The problems treated in the Manual have been collected during several IAEA missions in developing countries. 125 refs, 161 figs, 22 tabs

  4. Neutron monitor generated data distributions in quantum variational Monte Carlo

    Science.gov (United States)

    Kussainov, A. S.; Pya, N.

    2016-08-01

    We have assessed the potential applications of the neutron monitor hardware as random number generator for normal and uniform distributions. The data tables from the acquisition channels with no extreme changes in the signal level were chosen as the retrospective model. The stochastic component was extracted by fitting the raw data with splines and then subtracting the fit. Scaling the extracted data to zero mean and variance of one is sufficient to obtain a stable standard normal random variate. Distributions under consideration pass all available normality tests. Inverse transform sampling is suggested to use as a source of the uniform random numbers. Variational Monte Carlo method for quantum harmonic oscillator was used to test the quality of our random numbers. If the data delivery rate is of importance and the conventional one minute resolution neutron count is insufficient, we could always settle for an efficient seed generator to feed into the faster algorithmic random number generator or create a buffer.

  5. Current Status and Progress of Developing a D-D Neutron Generator

    International Nuclear Information System (INIS)

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

    2006-01-01

    The research to develop a D-D neutron generator was begun in 2001. A prototype device was built in 2004, and partly modified in 2005. By using the modified prototype D-D neutron generator, neutron generation runs were performed, and the characteristics of D-D neutron generation was investigated. The final goal of maximum neutron yield is 10 8 n/s, while a yield of 6.5x10 7 n/s has been achieved. Here, the results of neutron generation runs performed by using the modified prototype device are summarized, and the feature of a new ion source to be tested in weeks is briefly described

  6. A Proposal for a Next Generation European Neutron Source

    International Nuclear Information System (INIS)

    Andersen, K.H.; Carlile, C.J.

    2016-01-01

    We argue that it is not too early to begin the planning process for a next generation neutron source for Europe, even as the European Spallation Source is being constructed. We put forward three main arguments. Firstly, nowadays the period between the first scientific concept of a new facility being proposed and its actual realisation is approaching half a century. We show evidence for this. Secondly, there is a straightforward development of the short pulse/long pulse spallation concepts that will deliver gains in neutron brightness of more than a factor 30 over what the ESS will soon deliver and provide the optimum balance between resolution and intensity. We describe our concept, which is a spallation source where the proton pulse length is matched to the moderating time of slow neutrons. Thirdly, when we look at our colleagues in astronomy and high energy physics, we see that they have a totally different, more global and more ambitious approach to the coming generations of large facilities. We argue that it is time for the neutron community not simply to rest upon its laurels and take what is given but to be proactive.. (paper)

  7. The first IEC fusion industrial neutron generator and developments

    Science.gov (United States)

    Sved, John

    1999-06-01

    Inertial Electrostatic Confinement fusion grade plasma containment has been sporadically researched since the early 1960's. In the 1990's the work of G. H. Miley and his team at the University of Illinios, Fusion Studies Laboratory, Champaign-Urbana has stimulated a collaboration with industry. The development and test program for the first industrial IEC neutron generator has progressed to the point where an endurance test is under way to demonstrate at least 10,000 hours of operational life of the sealed chamber device without servicing. The market entry goals of steady 107 D-D n/s CW output with an air-cooled system have been achieved. DASA has invested in the development of the industrial product and the continuing basic research at the UI-FSL. The complete DASA FusionStar IEC-PS1 point source neutron generator set is described with emphasis on the interfaces to user NAA systems. The next product developments are pulsed neutron operations and higher fusion reaction rates of up to 1010 by means of affordable add-ons to the basic IEC-PS system. The production engineering experience gained will next be applied to a more challenging line source variant of the IEC. Beyond neutron and proton sources, several other IEC applications are being developed.

  8. Shielding evaluation of neutron generator hall by Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

    2017-04-01

    A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

  9. Superpower proton linear accelerators for neutron generators and electronuclear facilities

    International Nuclear Information System (INIS)

    Lazarev, N.V.; Kozodaev, A.M.

    2000-01-01

    The report is a review of projects on the superpower proton linear accelerators (SPLA) for neutron generators (NG) and electronuclear facilities, proposed in the recent years. The beam average output capacity in these projects reaches 100 MW. The basic parameters of certain operating NGs, as well as some projected NGs will the SPLA drivers are presented. The problems on application of superconducting resonators in the SPLA as well as the issues of the SPLA reliability and costs are discussed [ru

  10. A liquid hydrocarbon deuteron source for neutron generators

    Science.gov (United States)

    Schwoebel, P. R.

    2017-06-01

    Experimental studies of a deuteron spark source for neutron generators using hydrogen isotope fusion reactions are reported. The ion source uses a spark discharge between electrodes coated with a deuterated hydrocarbon liquid, here Santovac 5, to inhibit permanent electrode erosion and extend the lifetime of high-output neutron generator spark ion sources. Thompson parabola mass spectra show that principally hydrogen and deuterium ions are extracted from the ion source. Hydrogen is the chief residual gas phase species produced due to source operation in a stainless-steel vacuum chamber. The prominent features of the optical emission spectra of the discharge are C+ lines, the hydrogen Balmer Hα-line, and the C2 Swan bands. Operation of the ion source was studied in a conventional laboratory neutron generator. The source delivered an average deuteron current of ˜0.5 A nominal to the target in a 5 μs duration pulse at 1 Hz with target voltages of -80 to -100 kV. The thickness of the hydrocarbon liquid in the spark gap and the consistency thereof from spark to spark influences the deuteron yield and plays a role in determining the beam-focusing characteristics through the applied voltage necessary to break down the spark gap. Higher breakdown voltages result in larger ion beam spots on the target and vice-versa. Because the liquid self-heals and thereby inhibits permanent electrode erosion, the liquid-based source provides long life, with 104 pulses to date, and without clear evidence that, in principle, the lifetime could not be much longer. Initial experiments suggest that an alternative cylindrical target-type generator design can extract approximately 10 times the deuteron current from the source. Preliminary data using the deuterated source liquid as a neutron-producing target are also presented.

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

    International Nuclear Information System (INIS)

    Shypailo, R.J.; Ellis, K.J.

    2008-01-01

    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 14.2 MeV neutrons. To compare the performance of these two units in our present PGNA system, we performed Monte Carlo simulations (MCNP-5; Los Alamos National Laboratory) evaluating the nitrogen reactions produced in tissue-equivalent phantoms and the effects of background interference on the gamma-detectors. Monte Carlo response curves showed increased gamma production per unit dose when using the D-D generator, suggesting that it is the more suitable choice for smaller sized subjects. The increased penetration by higher energy neutrons produced by the D-T generator supports its utility when examining larger, especially obese, subjects. A clinical PGNA analysis design incorporating both neutron generator options may be the best choice for a system required to measure a wide range of subject phenotypes. (author)

  12. Response of a neutron monitor area with TLDs pairs

    Energy Technology Data Exchange (ETDEWEB)

    Guzman G, K. A.; Borja H, C. G.; Valero L, C.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Gallego, E.; Lorente, A., E-mail: ing_karen_guzman@yahoo.com.mx [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

    2011-10-15

    The response of a passive neutron monitor area has been calculated using the Monte Carlo code MCNP5. The response was the amount of n({sup 6}Li, T){alpha} reactions occurring in a TLD-600 located at the center of a cylindrical polyethylene moderator. Fluence, (n, a) and H*(10) responses were calculated for 47 monoenergetic neutron sources. The H*(10) relative response was compared with responses of commercially available neutron monitors being alike. Due to {sup 6}Li cross section (n, {alpha}) reactions are mainly produced by thermal neutrons, however TLD-600 is sensitive to gamma-rays; to eliminate the signal due to photons monitor area was built to hold 2 pairs of TLD-600 and 2 pairs of TLD-700, thus from the difference between TLD-600 and TLD-700 readouts the net signal due to neutrons is obtained. The monitor area was calibrated at the Universidad Politecnica de Madrid using a {sup 241}AmBe neutron source; net TLD readout was compared with the H*(10) measured with a Bert hold Lb-6411. Performance of the neutron monitor area was determined through two independent experiments, in both cases the H*(10) was statistically equal to H*(10) measured with a Bert hold Lb-6411. Neutron monitor area with TLDs pairs can be used in working areas with intense, mixed and pulsed radiation fields. (Author)

  13. SPECTRUM WEIGHTED RESPONSES OF SEVERAL DETECTORS IN MIXED FIELDS OF FAST AND THERMAL NEUTRONS

    Directory of Open Access Journals (Sweden)

    SANG IN KIM

    2014-04-01

    Full Text Available The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software ‘K-SWR’. The detectors’ response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403. The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the 241Am-Be sources held in a graphite pile, a bare 241Am-Be source, and a DT neutron generator. Fluence-average energy (Eave varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [H*(10/h] varied from 0.99 to 16.5 mSv/h.

  14. Review of current neutron detection systems for emergency response

    Science.gov (United States)

    Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul; Kruschwitz, Craig

    2014-09-01

    Neutron detectors are used in a myriad of applications—from safeguarding special nuclear materials (SNM) to determining lattice spacing in soft materials. The transformational changes taking place in neutron detection and imaging techniques in the last few years are largely being driven by the global shortage of helium-3 (3He). This article reviews the status of neutron sensors used specifically for SNM detection in radiological emergency response. These neutron detectors must be highly efficient, be rugged, have fast electronics to measure neutron multiplicity, and be capable of measuring direction of the neutron sources and possibly image them with high spatial resolution. Neutron detection is an indirect physical process: neutrons react with nuclei in materials to initiate the release of one or more charged particles that produce electric signals that can be processed by the detection system. Therefore, neutron detection requires conversion materials as active elements of the detection system; these materials may include boron-10 (10B), lithium-6 (6Li), and gadollinium-157 (157Gd), to name a few, but the number of materials available for neutron detection is limited. However, in recent years, pulse-shape-discriminating plastic scintillators, scintillators made of helium-4 (4He) under high pressure, pillar and trench semiconductor diodes, and exotic semiconductor neutron detectors made from uranium oxide and other materials have widely expanded the parameter space in neutron detection methodology. In this article we will pay special attention to semiconductor-based neutron sensors. Modern microfabricated nanotubes covered inside with neutron converter materials and with very high aspect ratios for better charge transport will be discussed.

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

    Science.gov (United States)

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

    2014-12-01

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

  16. The CLYC-6 and CLYC-7 response to γ-rays, fast and thermal neutrons

    International Nuclear Information System (INIS)

    Giaz, A.; Pellegri, L.; Camera, F.; Blasi, N.; Brambilla, S.; Ceruti, S.; Million, B.; Riboldi, S.; Cazzaniga, C.; Gorini, G.; Nocente, M.; Pietropaolo, A.; Pillon, M.; Rebai, M.; Tardocchi, M.

    2016-01-01

    The crystal Cs 2 LiYCl 6 :Ce (CLYC) is a very interesting scintillator material because of its good energy resolution and its capability to identify γ-rays and fast/thermal neutrons. The crystal Cs 2 LiYCl 6 :Ce contains 6 Li and 35 Cl isotopes, therefore, it is possible to detect thermal neutrons through the reaction 6 Li(n, α)t while 35 Cl ions allow to measure fast neutrons through the reactions 35 Cl(n, p) 35 S and 35 Cl(n, α) 32 P. In this work two CLYC 1″×1″ crystals were used: the first crystal, enriched with 6 Li at 95% (CLYC-6) is ideal for thermal neutron measurements while the second one, enriched with 7 Li at >99% (CLYC-7) is suitable for fast neutron measurements. The response of CLYC scintillators was measured with different PMT models: timing or spectroscopic, with borosilicate glass or quartz window. The energy resolution, the neutron-γ discrimination and the internal activity are discussed. The capability of CLYC scintillators to discriminate γ rays from neutrons was tested with both thermal and fast neutrons. The thermal neutrons were measured with both detectors, using an AmBe source. The measurements of fast neutrons were performed at the Frascati Neutron Generator facility (Italy) where a deuterium beam was accelerated on a deuterium or on a tritium target, providing neutrons of 2.5 MeV or 14.1 MeV, respectively. The different sensitivity to thermal and fast neutrons of a CLYC-6 and of a CLYC-7 was additionally studied.

  17. The neutrons generator becomes miniature; Le generateur de neutrons se miniaturise

    Energy Technology Data Exchange (ETDEWEB)

    Boulben, A

    2005-05-01

    A mini neutrons generator has been designed by researchers of the California university of Los Angeles which does not need any high voltage power supply ar any radioactive element. The system comprises a tubular vacuum chamber of few tenths of cm filled up with gaseous deuterium at very low pressure (0.7 Pa) and an erbium deuteride (ErD{sub 3}) screen. A lithium tantalate (LiTaO{sub 3}) pyroelectric crystal and a tungsten electrode are used to ionize and accelerate the gaseous deuterium with a minimum energy need. The collision of deuterium ions with the screen generates a maximum flow rate of about a thousand neutrons per second. Short paper. (J.S.)

  18. Plastic fiber scintillator response to fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Danly, C. R.; Sjue, S.; Wilde, C. H.; Merrill, F. E.; Haight, R. C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)

    2014-11-15

    The Neutron Imaging System at NIF uses an array of plastic scintillator fibers in conjunction with a time-gated imaging system to form an image of the neutron emission from the imploded capsule. By gating on neutrons that have scattered from the 14.1 MeV DT energy to lower energy ranges, an image of the dense, cold fuel around the hotspot is also obtained. An unmoderated spallation neutron beamline at the Weapons Neutron Research facility at Los Alamos was used in conjunction with a time-gated imaging system to measure the yield of a scintillating fiber array over several energy bands ranging from 1 to 15 MeV. The results and comparison to simulation are presented.

  19. Sound response of superheated drop bubble detectors to neutrons

    International Nuclear Information System (INIS)

    Gao Size; Chen Zhe; Liu Chao; Ni Bangfa; Zhang Guiying; Zhao Changfa; Xiao Caijin; Liu Cunxiong; Nie Peng; Guan Yongjing

    2012-01-01

    The sound response of the bubble detectors to neutrons by using 252 Cf neutron source was described. Sound signals were filtered by sound card and PC. The short-time signal energy. FFT spectrum, power spectrum, and decay time constant were got to determine the authenticity of sound signal for bubbles. (authors)

  20. Evaluation of response matrix of a multisphere neutron spectrometer ...

    Indian Academy of Sciences (India)

    Abstract. Neutron energy responses of water sphere spectrometers (WSS) to 30 MeV have been calculated by means of Monte Carlo calculations, using the computer code MCNP4C with ENDF/. B-VI.0 neutron cross-section. The calculations have been performed for 3He detector (typical SP9) placed inside 2, 3, 5, 8, ...

  1. An intense neutron generator based on a proton accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Bartholomew, G A; Milton, J C.D.; Vogt, E W

    1964-07-01

    A study has been made of the demand for a neutron facility with a thermal flux of {>=} 10{sup 16} n cm{sup -2} sec{sup -1} and of possible methods of producing such fluxes with existing or presently developing technology. Experimental projects proposed by neutron users requiring high fluxes call for neutrons of all energies from thermal to 100 MeV with both continuous-wave and pulsed output. Consideration of the heat generated in the source per useful neutron liberated shows that the (p,xn) reaction with 400 1000 MeV bombarding energies and heavy element targets (e.g. bismuth, lead) is capable of greater specific source strength than other possible methods realizable within the time scale. A preliminary parameter optimization carried through for the accelerator currently promising greatest economy (the separated orbit cyclotron or S.O.C.), reveals that a facility delivering a proton beam of about 65 mA at about 1 BeV would satisfy the flux requirement with a neutron cost significantly more favourable than that projected for a high flux reactor. It is suggested that a proton storage ring providing post-acceleration pulsing of the proton beam should be developed for the facility. With this elaboration, and by taking advantage of the intrinsic microscopic pulse structure provided by the radio frequency duty cycle, a very versatile source may be devised capable of producing multiple beams of continuous and pulsed neutrons with a wide range of energies and pulse widths. The source promises to be of great value for high flux irradiations and as a pilot facility for advanced reactor technology. The proposed proton accelerator also constitutes a meson source capable of producing beams of {pi} and {mu} mesons and of neutrinos orders of magnitude more intense than those of any accelerator presently in use. These beams, which can be produced simultaneously with the neutron beams, open vast areas of new research in fundamental nuclear structure, elementary particle physics

  2. An intense neutron generator based on a proton accelerator

    International Nuclear Information System (INIS)

    Bartholomew, G.A.; Milton, J.C.D.; Vogt, E.W.

    1964-01-01

    A study has been made of the demand for a neutron facility with a thermal flux of ≥ 10 16 n cm -2 sec -1 and of possible methods of producing such fluxes with existing or presently developing technology. Experimental projects proposed by neutron users requiring high fluxes call for neutrons of all energies from thermal to 100 MeV with both continuous-wave and pulsed output. Consideration of the heat generated in the source per useful neutron liberated shows that the (p,xn) reaction with 400 1000 MeV bombarding energies and heavy element targets (e.g. bismuth, lead) is capable of greater specific source strength than other possible methods realizable within the time scale. A preliminary parameter optimization carried through for the accelerator currently promising greatest economy (the separated orbit cyclotron or S.O.C.), reveals that a facility delivering a proton beam of about 65 mA at about 1 BeV would satisfy the flux requirement with a neutron cost significantly more favourable than that projected for a high flux reactor. It is suggested that a proton storage ring providing post-acceleration pulsing of the proton beam should be developed for the facility. With this elaboration, and by taking advantage of the intrinsic microscopic pulse structure provided by the radio frequency duty cycle, a very versatile source may be devised capable of producing multiple beams of continuous and pulsed neutrons with a wide range of energies and pulse widths. The source promises to be of great value for high flux irradiations and as a pilot facility for advanced reactor technology. The proposed proton accelerator also constitutes a meson source capable of producing beams of π and μ mesons and of neutrinos orders of magnitude more intense than those of any accelerator presently in use. These beams, which can be produced simultaneously with the neutron beams, open vast areas of new research in fundamental nuclear structure, elementary particle physics, and perhaps also in

  3. Timing reference generators and chopper controllers for neutron sources

    International Nuclear Information System (INIS)

    Nelson, R.; Merl, R.; Rose, C.

    2001-01-01

    Due to AC-power-grid frequency fluctuations, the designers for accelerator-based spallation-neutron facilities have worked to optimize the competing and contrasting demands of accelerator and neutron chopper performance. Powerful new simulation techniques have enabled the modeling of the timing systems that integrate chopper controllers and chopper hardware. For the first time, we are able to quantitatively access the tradeoffs between these two constraints and design or upgrade a facility to optimize total system performance. Thus, at LANSCE, we now operate multiple chopper systems and the accelerator as simple slaves to a single master-timing-reference generator. For the SNS we recommend a similar system that is somewhat less tightly coupled to the power grid. (author)

  4. Activation analysis with neutron generators using short-lived radionuclides

    International Nuclear Information System (INIS)

    Salma, I.

    1993-01-01

    The short half-life involves a number of important differences in production, transportation and measurement of radionuclides, and in counting statistics as compared with those in traditional activation analysis. Experiments were performed to investigate the analytical possibilities and prospective utilization of short-lived radionuclides produced by 14-MeV neutron irradiation. A rapid pneumatic transfer system for use with neutron generators was installed and applied for detecting radionuclides with a half-life from 300 ms to 30 s. The transport time for samples with a total mass of 1-4 g is between 130 and 160 ms for pressurized air of 0.1-0.4 MPa. 11 elements were studied by the conventional activation method using both a typical pneumatic transport system (run time 3 s) and the fast pneumatic transport facility. The effect of the cyclic activation technique on the elemental sensitivities was also investigated. (orig.)

  5. Fast Neutrons - LET Distributions and the Response of Mammalian Cells

    Energy Technology Data Exchange (ETDEWEB)

    Bewley, D. K. [Medical Research Council Cyclotron Unit, Hammersmith Hospital, London (United Kingdom)

    1968-03-15

    Distributions of stopping power (LET) are given for four beams of fast neutrons, namely, neutrons of 14.6 MeV, neutrons produced by bombarding a thick beryllium target with 15 MeV deuterons, neutrons of 3 MeV, and fast neutrons produced by bombardment of a {sup 235}U converter plate with thermal neutrons. The track average LET is correlated with mean neutron energy, but the dose average is approximately constant. However, neither of these types of average is expected to have much relevance to radiobiology. Further, specification of a ''biologically effective LET'' depends on the biological test used, and is not solely a function of the radiation quality. An attempt has been made to calculate the response of T.I. kidney cells in tissue culture to these four beams of neutrons, based on their response to charged particles using the track-segment method. The calculated RBE's of the neutron beams are lower than the observed values and the calculated values of the oxygen enhancement ratio are higher. These differences seem too great to be explained by errors in dosimetry and in the calculated LET spectra. The suggestion is made that LET is not an adequate criterion of radiation quality, and that the discrepancies may be explained by more detailed consideration of the part played by delta rays and by heavy recoil tracks of short range. (author)

  6. Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

    Science.gov (United States)

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

    2017-04-01

    Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

  7. Neutron Generators Developed at LBNL for Homeland Security and Imaging Applications

    International Nuclear Information System (INIS)

    Reijonen, Jani

    2006-01-01

    The Plasma and Ion Source Technology Group at Lawrence Berkeley National Laboratory has developed various types of advanced D-D (neutron energy 2.5 MeV), D-T (14 MeV) and T-T (0-9 MeV) neutron generators for wide range of applications. These applications include medical (Boron Neutron Capture Therapy), homeland security (Prompt Gamma Activation Analysis, Fast Neutron Activation Analysis and Pulsed Fast Neutron Transmission Spectroscopy) and planetary exploration with a sub-surface material characterization on Mars. These neutron generators utilize RF induction discharge to ionize the deuterium/tritium gas. This discharge method provides high plasma density for high output current, high atomic species from molecular gases, long life operation and versatility for various discharge chamber geometries. Four main neutron generator developments are discussed here: high neutron output co-axial neutron generator for BNCT applications, point neutron generator for security applications, compact and sub-compact axial neutron generator for elemental analysis applications. Current status of the neutron generator development with experimental data will be presented

  8. Compact Neutron Generators for Medical Home Land Security and Planetary Exploration

    International Nuclear Information System (INIS)

    Reijonen, J.

    2005-01-01

    The Plasma and Ion Source Technology Group at Lawrence Berkeley National Laboratory has developed various types of advanced D-D (neutron energy 2.5 MeV), D-T (14 MeV) and T-T (0-9 MeV) neutron generators for wide range of applications. These applications include medical (Boron Neutron Capture Therapy), homeland security (Prompt Gamma Activation Analysis, Fast Neutron Activation Analysis and Pulsed Fast Neutron Transmission Spectroscopy) and planetary exploration with a sub-surface material characterization on Mars. These neutron generators utilize RF induction discharge to ionize the deuterium/tritium gas. This discharge method provides high plasma density for high output current, high atomic species from molecular gases, long life operation and versatility for various discharge chamber geometries. Four main neutron generator developments are discussed here: high neutron output co-axial neutron generator for BNCT applications, point neutron generator for security applications, compact and sub-compact axial neutron generator for elemental analysis applications. Current status of the neutron generator development with experimental data will be presented

  9. Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, P., E-mail: peter.andersson@physics.uu.se; Andersson-Sunden, E.; Sjöstrand, H.; Jacobsson-Svärd, S. [Department of Physics and Astronomy, Division of Applied Nuclear Physics, Uppsala University, Lägerhyddsgatan 1, 751 20 Uppsala (Sweden)

    2014-08-01

    In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm{sup −1}, solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful

  10. Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator.

    Science.gov (United States)

    Andersson, P; Andersson-Sunden, E; Sjöstrand, H; Jacobsson-Svärd, S

    2014-08-01

    In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm(-1), solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful

  11. Materials-based process tolerances for neutron generator encapsulation

    International Nuclear Information System (INIS)

    Berry, Ryan S.; Adolf, Douglas Brian; Stavig, Mark Edwin

    2007-01-01

    Variations in the neutron generator encapsulation process can affect functionality. However, instead of following the historical path in which the effects of process variations are assessed directly through functional tests, this study examines how material properties key to generator functionality correlate with process variations. The results of this type of investigation will be applicable to all generators and can provide insight on the most profitable paths to process and material improvements. Surprisingly, the results at this point imply that the process is quite robust, and many of the current process tolerances are perhaps overly restrictive. The good news lies in the fact that our current process ensures reproducible material properties. The bad new lies in the fact that it would be difficult to solve functional problems by changes in the process

  12. Materials-based process tolerances for neutron generator encapsulation.

    Energy Technology Data Exchange (ETDEWEB)

    Berry, Ryan S.; Adolf, Douglas Brian; Stavig, Mark Edwin

    2007-10-01

    Variations in the neutron generator encapsulation process can affect functionality. However, instead of following the historical path in which the effects of process variations are assessed directly through functional tests, this study examines how material properties key to generator functionality correlate with process variations. The results of this type of investigation will be applicable to all generators and can provide insight on the most profitable paths to process and material improvements. Surprisingly, the results at this point imply that the process is quite robust, and many of the current process tolerances are perhaps overly restrictive. The good news lies in the fact that our current process ensures reproducible material properties. The bad new lies in the fact that it would be difficult to solve functional problems by changes in the process.

  13. Proposal for an accelerator-based neutron generator

    International Nuclear Information System (INIS)

    Grand, P.

    1975-07-01

    An Accelerator-based Neutron Generator is described that consists of a 30-MeV deuteron linear accelerator using a flowing liquid lithium target. With a continuous deuteron current of 100 milliamperes, a source intensity of more than 10 16 neutrons per second will be produced. The neutrons will be emitted in a roughly collimated beam. The proposed facility can be divided into two areas: the 30-MeV linear accelerator and the multiple-target experimental area. The 30-MeV accelerator will consist of eight rf accelerating cavities in a single vacuum tank, each cavity being powered by its own rf power amplifier operating at 50 MHz. To shield the beam bunches from the rf field when it is in the decelerating direction, 66 ''drift tubes'' will be included; the drift-tube structures will include quadrupole magnets which will keep the beam focused. The accelerator will produce a continuous beam of 100 milliamperes. Beam power will thus be 3.0 megawatts; total power including rf losses in the accelerating cavities will be 4.5 megawatts. The injectors for the linear accelerator will be two 500-kV dc accelerators, one for injection of D + ions and the other for D - ions. They can be used simultaneously or one can serve as a spare in case of breakdown or maintenance of the other. (U.S.)

  14. A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

    1999-01-01

    In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

  15. Monte Carlo Simulation on Compensated Neutron Porosity Logging in LWD With D-T Pulsed Neutron Generator

    International Nuclear Information System (INIS)

    Zhang Feng; Hou Shuang; Jin Xiuyun

    2010-01-01

    The process of neutron interaction induced by D-T pulsed neutron generator and 241 Am-Be source was simulated by using Monte Carlo method. It is concluded that the thermal neutron count descend exponentially as the spacing increasing. The smaller porosity was, the smaller the differences between the two sources were. When the porosity reached 40%, the ratio of thermal neutron count generated by D-T pulsed neutron source was much larger than that generated by 241 Am-Be neutron source, and its distribution range was wider. The near spacing selected was 20-30 cm, and that of far spacing was about 60-70 cm. The detection depth by using D-T pulsed neutron source was almost unchanged under condition of the same sapcing, and the sensitivity of measurement to the formation porosity decreases. The results showed that it can not only guarantee the statistic of count, but also improve detection sensitivity and depth at the same time of increasing spacing. Therefore, 241 Am-Be neutron source can be replaced by D-T neutron tube in LWD tool. (authors)

  16. Trial production of hyper-thermal neutron generator for Neutron Capture Therapy (NCT) and its radiation properties

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Toru

    1999-01-01

    In NCT, it was at first important to give a cancer portion to radiation dose required for its recovery. By finding out that whole cross-section of water comprising of a living body decreased monotonously with increase of neutron energy from about 100 barn against thermal neutron, became about 40 barn at about 0.5 eV and kept constant to 40 barn till at about 100 eV, application of thermal neutron shifted to higher temperature side, called Hyper thermal neutron, to NCT is proposed. The Hyper thermal neutron radiation can be expected to have similar controllability to that of the thermal neutron radiation. In 1977 fiscal year, a trial Hyper thermal neutron generator was produced on a base of up-to-date investigation results. As a part of property evaluation of the generator, evaluation of energy spectra in the Hyper thermal neutron generated at LINAC by TOF was conducted to confirm shift of the spectra to high temperature side. And, a Fantom experiment at KUR heavy water neutron radiation facility was also conducted to confirm effect of improvement in deep portion dose distribution. (G.K.)

  17. Evaluation of response function of moderating-type neutron detector and application to environmental neutron measurement

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakamura, Takashi; Iwai, Satoshi; Katsuki, Shinji; Kamata, Masashi.

    1983-08-01

    The energy-dependent response function of a multi-cylinder moderating-type BF 3 counter, so-called Bonner counter, was calculated by the time-dependent multi-group Monte Carlo code, TMMCR. The calculated response function was evaluated experimentally for neutron energy below about 50 keV down to epithermal energy by the time-of-flight method combining with a large lead pile at the Nuclear Engineering Research Laboratory, University of Tokyo and also above 50 keV by using the monoenergetic neutron standard field a t the Electrotechnical Laboratory. The time delay in the polyethylene moderator of the Bonner counter due to multiple collisions with hydrogen was analyzed by the TMMCR code and used for the time-spectrum analysis of the time-of-flight measurement. The response function obtained by these two experiments showed good agreement with the calculated results. This Bonner counter having a response function evaluated from thermal to MeV energy range was used for spectrometry and dosimetry of environmental neutrons around some nuclear facilities. The neutron spectra and dose measured in the environment around a 252 Cf fission source, fast neutron source reactor and electron synchrotron were all in good agreement with the calculated results and the measured results with other neutron detectors. (author)

  18. Neutron source investigations in support of the cross section program at the Argonne Fast-Neutron Generator

    International Nuclear Information System (INIS)

    Meadows, J.W.; Smith, D.L.

    1980-05-01

    Experimental methods related to the production of neutrons for cross section studies at the Argonne Fast-Neutron Generator are reviewed. Target assemblies commonly employed in these measurements are described, and some of the relevant physical properties of the neutron source reactions are discussed. Various measurements have been performed to ascertain knowledge about these source reaction that is required for cross section data analysis purposes. Some results from these studies are presented, and a few specific examples of neutron-source-related corrections to cross section data are provided. 16 figures, 3 tables

  19. Characterizing Scintillator Response with Neutron Time-of-Flight

    Science.gov (United States)

    Palmisano, Kevin; Visca, Hannah; Caves, Louis; Wilkinson, Corey; McClow, Hannah; Padalino, Stephen; Forrest, Chad; Katz, Joe; Sangster, Craig; Regan, Sean

    2017-10-01

    Neutron scintillator diagnostics for ICF can be characterized using the neutron time-of-flight (nTOF) line on Geneseo's 1.7 MV Tandem Pelletron Accelerator. Neutron signals can be differentiated from gamma signals by employing a coincidence method called the associated particle technique (APT). In this measurement, a 2.1 MeV beam of deuterons incident on a deuterated polyethylene target produces neutrons via the d(d,n)3He reaction. A BC-412 plastic scintillator, placed at a scattering angle of 152º, detects 1.76 MeV neutrons in coincidence with the 2.56 MeV 3He ions at an associated angle of 10º. The APT is used to identify the 1.76 MeV neutron while the nTOF line determines its energy. By gating only mono-energetic neutrons, the instrument response function of the scintillator can be determined free from background scattered neutrons and gamma rays. Funded in part by a Grant from the DOE, through the Laboratory for Laser Energetics.

  20. First PGAA and NAA experimental results from a compact high intensity D-D neutron generator

    International Nuclear Information System (INIS)

    Reijonen, J.; Leung, K.-N.; Firestone, R.B.; English, J.A.; Perry, D.L.; Smith, A.; Gicquel, F.; Sun, M.; Bandong, B.; Garabedian, G.; Revay, Zs.; Szentmiklosi, L.; Molnar, G.

    2003-01-01

    Various types of neutron generator systems have been designed and tested at the Plasma and Ion Source Technology Group at Lawrence Berkeley National Laboratory. These generators are based on a D-D fusion reaction. These high power D-D neutron generators can provide neutron fluxes in excess of the current state of the art D-T neutron generators, without the use of pre-loaded targets or radioactive tritium gas. Safe and reliable long-life operations are the typical features of these D-D generators. All of the neutron generators developed in the Plasma and Ion Source Technology Group are utilizing powerful RF-induction discharge to generate the deuterium plasma. One of the advantages of using the RF-induction discharge is it's ability to generate high fraction of atomic ions from molecular gases, and the ability to generate high plasma densities for high extractable ion current from relatively small discharge volume

  1. In phantom dosimetric response of tooth enamel to neutrons

    International Nuclear Information System (INIS)

    Fattibene, P.; Angelone, M.; Pillon, M.; De Coste, V.

    2004-01-01

    Electron Paramagnetic Resonance dosimetry based on tooth enamel has one important application in dose reconstruction of nuclear plant workers, where the contribution of neutrons to individual dose is often important. Evaluation of tooth enamel response to neutrons is thus an important goal. A few experimental data at thermal and fast neutron energies are available. A first evaluation of the tooth enamel relative response to 60 Co in monoenergetic neutron flux of 2.8 and of 14 MeV, published elsewhere, has provided results apparently non-consistent with the results obtained at lower and higher energies. A comparison of those results in the 2.8 and 14 MeV beams with those available in the literature for other beams is reported and possible reasons for incongruities are discussed. Dose conversion factors of enamel to the water and air are also calculated and reported. (authors)

  2. Compact D-D/D-T neutron generators and their applications

    Energy Technology Data Exchange (ETDEWEB)

    Lou, Tak Pui [Univ. of California, Berkeley, CA (United States)

    2003-01-01

    Neutron generators based on the 2H(d,n)3He and 3H(d,n)4He fusion reactions are the most commonly available neutron sources. The applications of current commercial neutron generators are often limited by their low neutron yield and their short operational lifetime. A new generation of D-D/D-T fusion-based neutron generators has been designed at Lawrence Berkeley National Laboratory (LBNL) by using high current ion beams hitting on a self-loading target that has a large surface area to dissipate the heat load. This thesis describes the rationale behind the new designs and their potential applications. A survey of other neutron sources is presented to show their advantages and disadvantages compared to the fusion-based neutron generator. A prototype neutron facility was built at LBNL to test these neutron generators. High current ion beams were extracted from an RF-driven ion source to produce neutrons. With an average deuteron beam current of 24 mA and an energy of 100 keV, a neutron yield of >109 n/s has been obtained with a D-D coaxial neutron source. Several potential applications were investigated by using computer simulations. The computer code used for simulations and the variance reduction techniques employed were discussed. A study was carried out to determine the neutron flux and resolution of a D-T neutron source in thermal neutron scattering applications for condensed matter experiments. An error analysis was performed to validate the scheme used to predict the resolution. With a D-T neutron yield of 1014 n/s, the thermal neutron flux at the sample was predicted to be 7.3 x 105 n/cm2s. It was found that the resolution of cold neutrons was better than that of thermal neutrons when the duty factor is high. This neutron generator could be efficiently used for research and educational purposes at universities. Additional applications studied were positron production and

  3. Compact D-D/D-T neutron generators and their applications

    International Nuclear Information System (INIS)

    Lou, Tak Pui

    2003-01-01

    Neutron generators based on the 2 H(d,n) 3 He and 3 H(d,n) 4 He fusion reactions are the most commonly available neutron sources. The applications of current commercial neutron generators are often limited by their low neutron yield and their short operational lifetime. A new generation of D-D/D-T fusion-based neutron generators has been designed at Lawrence Berkeley National Laboratory (LBNL) by using high current ion beams hitting on a self-loading target that has a large surface area to dissipate the heat load. This thesis describes the rationale behind the new designs and their potential applications. A survey of other neutron sources is presented to show their advantages and disadvantages compared to the fusion-based neutron generator. A prototype neutron facility was built at LBNL to test these neutron generators. High current ion beams were extracted from an RF-driven ion source to produce neutrons. With an average deuteron beam current of 24 mA and an energy of 100 keV, a neutron yield of >10 9 n/s has been obtained with a D-D coaxial neutron source. Several potential applications were investigated by using computer simulations. The computer code used for simulations and the variance reduction techniques employed were discussed. A study was carried out to determine the neutron flux and resolution of a D-T neutron source in thermal neutron scattering applications for condensed matter experiments. An error analysis was performed to validate the scheme used to predict the resolution. With a D-T neutron yield of 10 14 n/s, the thermal neutron flux at the sample was predicted to be 7.3 x 10 5 n/cm 2 s. It was found that the resolution of cold neutrons was better than that of thermal neutrons when the duty factor is high. This neutron generator could be efficiently used for research and educational purposes at universities. Additional applications studied were positron production and Boron Neutron Capture Therapy (BNCT). The neutron flux required for positron

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. The space distribution of neutrons generated in massive lead target by relativistic nuclear beam

    International Nuclear Information System (INIS)

    Chultem, D.; Damdinsuren, Ts.; Enkh-Gin, L.; Lomova, L.; Perelygin, V.; Tolstov, K.

    1993-01-01

    The present paper is devoted to implementation of solid state nuclear track detectors in the research of the neutron generation in extended lead spallation target. Measured neutrons space distribution inside the lead target and neutron distribution in the thick water moderator are assessed. (Author)

  6. Optic fibber data acquisition and transmission system dedicated to a neutron generator

    International Nuclear Information System (INIS)

    Ledo Pereda, Luis Miguel; Vergara Limon, Sergio; Arteche Diaz, Raul

    2009-01-01

    Hereby, are presented the design, construction and application of a virtual data acquisition system based on the usage of microcontrollers, optic fibber, and PC. System is aimed to the reestablishment of the communication between the basic modules of a Neutron Generator. The work shows, how the original interface design is upgraded by the automation of the data acquisition, on the Neutron Generator exploitation parameters. The PC usage is being introduced in the Neutron Generator and the precedent is established for further subsystem

  7. Transient response of self-powered neutron detectors

    International Nuclear Information System (INIS)

    Boeck, H.; Gebureck, P.; Stegemann, D.

    The behaviour of self-powered neutron detectors with Co, Er, Hf and Pt emitters was investigated during reactor square wave and pulse operation. The detector's response was compared with the current of an excore ionization chamber. Characteristical deviations from linearity were observed with all detectors at fast reactor periods. The exact cause of these deviations is not yet fully understood but several possibilities for the nonlinear behaviour of self-powered neutron detectors are outlined. (author)

  8. Neutron generators with size scalability, ease of fabrication and multiple ion source functionalities

    Science.gov (United States)

    Elizondo-Decanini, Juan M

    2014-11-18

    A neutron generator is provided with a flat, rectilinear geometry and surface mounted metallizations. This construction provides scalability and ease of fabrication, and permits multiple ion source functionalities.

  9. Generation of neutron scattering cross sections for silicon dioxide

    International Nuclear Information System (INIS)

    Ramos, R; Marquez Damian, J.I; Granada, J.R.; Cantargi, F

    2009-01-01

    A set of neutron scattering cross sections for silicon and oxygen bound in silicon dioxide were generated and validated. The cross sections were generated in the ACE format for MCNP using the nuclear data processing system NJOY, and the validation was done with published experimental data. This cross section library was applied to the calculation of five critical configurations published in the benchmark Critical Experiments with Heterogeneous Compositions of Highly Enriched Uranium, Silicon Dioxide and Polyethylene. The original calculations did not use the thermal scattering libraries generated in this work and presented significant differences with the experimental results. For this reason, the newly generated library was added to the input and the multiplication factor for each configuration was recomputed. The utilization of the thermal scattering libraries did not result in an improvement of the computational results. Based on this we conclude that integral experiments to validate this type of thermal cross sections need to be designed with a higher influence of thermal scattering in the measured result, and the experiments have to be performed under more controlled conditions. [es

  10. Fusion neutron generation by high-repetitive target injection

    International Nuclear Information System (INIS)

    Kitagawa, Yoneyoshi

    2015-01-01

    Pellet injection and repetitive laser illumination are key technologies for realizing inertial fusion energy. The Graduate School for the Creation of New Photonics Industries, Hamamatsu Photonics K. K. and Toyota Motor Corporation demonstrate the pellet injection, counter laser beams' engagement and neutron generation. Deuterated polystyrene (CD) bead pellets, after free-falling for a distance of 18 cm at 1 Hz, are successfully engaged by two counter laser beams from a diode-pumped, ultra-intense laser HAMA. The laser energy, pulse duration, wavelength and the intensity are 0.63 J per beam, 104 fs, 811 nm and 4.7 x 10 18 W/cm 2 , respectively. The irradiated pellets produce D (D, n) 3 He-reacted neutrons with a maximum yield of 9.5 x 10 4 /4π sr/shot. A straight channel with 10 μm-diameter is found through the beads. The pellet size is 1 mm. The results indicate potentially useful technologies for the next step in realizing inertial fusion energy. The results are reviewed as well as some oversea activities. (author)

  11. Calculating the energy spectrum of neutrons from tritium target of the NG-150 type generator

    International Nuclear Information System (INIS)

    Bortash, A.I.; Kuznetsov, V.S.

    1987-01-01

    Calculation procedure of neutron spectra yielding from the NG-150 generator target chamber with regard to deutron moderation is suggested. Using the suggested procedure, neutron spectra for different escape angles formed in the tritium target are calculated. The spectrum of neutrons scattered in cooling water is calculated. The mean energy of neutrons escaping at the angle of 0 deg equalling 14.5 MeV is obtained

  12. Photon and neutron energy response of Thermoluminescent (TL) dosimeters

    International Nuclear Information System (INIS)

    Thilagam, L.; Priya, M.R.; Mohapatra, D.K.

    2018-01-01

    Theoretical Monte Carlo (MC) simulations are carried out to investigate the relative thermoluminesence (TL) response of the most commonly used TLD materials to a wide range of photon energy. The effect of polytetrafluoroethylene (PTFE) on TL response of CaSO 4 :Dy is also studied. Additionally, the neutron response of LiF:Mg,Ti TL materials with different concentrations of 6 Li is estimated in terms of the number of 6 Li(n, t) 4 He capture reactions for a wider neutron energy

  13. 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)

  14. Calibration Of A 14 MeV Neutron Generator With Reference To NBS-1

    International Nuclear Information System (INIS)

    Heimbach, Craig R.

    2011-01-01

    NBS-1 is the US national neutron reference source. It has a neutron emission rate (June 1961) of 1.257x10 6 n/s 1,2,3 with an uncertainty of 0.85%(k = 1). Neutron emission-rate calibrations performed at the National Institute of Standards and Technology (NIST) are made in comparison to this source, either directly or indirectly. To calibrate a commercial 14 MeV neutron generator, NIST performed a set of comparison measurements to evaluate the neutron output relative to NBS-1. The neutron output of the generator was determined with an uncertainty of about 7%(k = 1). The 15-hour half-life of one of the reactions used also makes possible off-site measurements. Consideration is given to similar calibrations for a 2.5 MeV neutron generator.

  15. Solid polystyrene and deuterated polystyrene light output response to fast neutrons

    International Nuclear Information System (INIS)

    Simpson, R.; Danly, C.; Merrill, F. E.; Volegov, P. L.; Wilde, C.; Glebov, V. Yu.; Hurlbut, C.

    2016-01-01

    The Neutron Imaging System has proven to be an important diagnostic in studying DT implosion characteristics at the National Ignition Facility. The current system depends on a polystyrene scintillating fiber array, which detects fusion neutrons born in the DT hotspot as well as neutrons that have scattered to lower energies in the surrounding cold fuel. Increasing neutron yields at NIF, as well as a desire to resolve three-dimensional information about the fuel assembly, have provided the impetus to build and install two additional next-generation neutron imaging systems. We are currently investigating a novel neutron imaging system that will utilize a deuterated polystyrene (CD) fiber array instead of standard hydrogen-based polystyrene (CH). Studies of deuterated xylene or deuterated benzene liquid scintillator show an improvement in imaging resolution by a factor of two [L. Disdier et al., Rev. Sci. Instrum. 75, 2134 (2004)], but also a reduction in light output [V. Bildstein et al., Nucl. Instrum. Methods Phys. Res., Sect. A 729, 188 (2013); M. I. Ojaruega, Ph.D. thesis, University of Michigan, 2009; M. T. Febbraro, Ph.D. thesis, University of Michigan, 2014] as compared to standard plastic. Tests of the relative light output of deuterated polystyrene and standard polystyrene were completed using 14 MeV fusion neutrons generated through implosions of deuterium-tritium filled capsules at the OMEGA laser facility. In addition, we collected data of the relative response of these two scintillators to a wide energy range of neutrons (1-800 MeV) at the Weapons Neutrons Research Facility. Results of these measurements are presented.

  16. Solid polystyrene and deuterated polystyrene light output response to fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, R., E-mail: raspberry@lanl.gov; Danly, C.; Merrill, F. E.; Volegov, P. L.; Wilde, C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Glebov, V. Yu. [Laboratory for Laser Energetics, Rochester, New York 14623 (United States); Hurlbut, C. [Eljen Technology, Sweetwater, Texas 79556 (United States)

    2016-04-15

    The Neutron Imaging System has proven to be an important diagnostic in studying DT implosion characteristics at the National Ignition Facility. The current system depends on a polystyrene scintillating fiber array, which detects fusion neutrons born in the DT hotspot as well as neutrons that have scattered to lower energies in the surrounding cold fuel. Increasing neutron yields at NIF, as well as a desire to resolve three-dimensional information about the fuel assembly, have provided the impetus to build and install two additional next-generation neutron imaging systems. We are currently investigating a novel neutron imaging system that will utilize a deuterated polystyrene (CD) fiber array instead of standard hydrogen-based polystyrene (CH). Studies of deuterated xylene or deuterated benzene liquid scintillator show an improvement in imaging resolution by a factor of two [L. Disdier et al., Rev. Sci. Instrum. 75, 2134 (2004)], but also a reduction in light output [V. Bildstein et al., Nucl. Instrum. Methods Phys. Res., Sect. A 729, 188 (2013); M. I. Ojaruega, Ph.D. thesis, University of Michigan, 2009; M. T. Febbraro, Ph.D. thesis, University of Michigan, 2014] as compared to standard plastic. Tests of the relative light output of deuterated polystyrene and standard polystyrene were completed using 14 MeV fusion neutrons generated through implosions of deuterium-tritium filled capsules at the OMEGA laser facility. In addition, we collected data of the relative response of these two scintillators to a wide energy range of neutrons (1-800 MeV) at the Weapons Neutrons Research Facility. Results of these measurements are presented.

  17. CIAE 600 kV ns pulse neutron generator

    International Nuclear Information System (INIS)

    Shen Guanren; Guan Xialing; Chen Hongtao

    2001-01-01

    The overall composition of CIAE 600 kV ns Pulse Neutron Generator (CPNG) are introduced, and its characteristic, main technological performance and application were also given. CPNG consists of high voltage power supply with highest output voltage 600 kV, direct current 15 mA, stability and ripple ≤0.1%, 2214 mm x 1604 mm x 1504 mm stainless steel high voltage electrode, built in head equipment uniform field accelerating tube, ns pulsed installation, turbomolecular vacuum pump system and drift pipes at 0 degree and 45 degree. Its characteristics are: (1) high current beam; (2) high current beam ns pulsed installation made use of low energy for chopper and high energy for buncher; (3) compactly laid out and simple in structure

  18. 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)

  19. 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)

  20. 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)

  1. Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

    Science.gov (United States)

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

    2014-12-01

    An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

    1994-11-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

  3. Fast neutron response of coumarin in water and heavy water

    International Nuclear Information System (INIS)

    Krishnan, D.; Kher, R.K.; Gopakumar, K.; Bhandari, N.S.

    1979-01-01

    Response of coumarin in aqueous solution has been studied earlier for gamma rays and fast neutrons by fluorescence measurement. For further fast neutron studies, two systems viz coumarin in H 2 0 and coumarin in D 2 0, were irradiated with fast neutrons in SNIF facility in the swimming pool type APSARA reactor at Trombay. Neutron fluence was estimated by measuring induced activity in sulphur pellet and associated gamma radiation was estimated using CaS0 4 :Dy TLD powder. The KERMA values were calculated for H 2 0 and D 2 0, assuming modified fission spectrum for fast neutron in SNIF position, and they were in the ratio of 2:1. Response of a chemical dosimetric system is expected to be proportional to the absorbed dose in the respective system for the same neutron fluence. This was experimentally found to be the case for coumarin in H 2 0 or D 2 0. These results are likely to be true in general for any aqueous chemical system. The limitations of using such a dual system for dosimetry in a mixed field is discussed. (author)

  4. Determination of the emission rate for the 14 MeV neutron generator with the use of radio-yttrium

    OpenAIRE

    Laszynska Ewa; Jednorog Slawomir; Ziolkowski Adam; Gierlik Michal; Rzadkiewicz Jacek

    2015-01-01

    The neutron emission rate is a crucial parameter for most of the radiation sources that emit neutrons. In the case of large fusion devices the determination of this parameter is necessary for a proper assessment of the power release and the prediction for the neutron budget. The 14 MeV neutron generator will be used for calibration of neutron diagnostics at JET and ITER facilities. The stability of the neutron generator working parameters like emission and angular homogeneity affects the accu...

  5. Analysis of the Neutron Generator and Target for the LSDTS System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Je; Lee, Yong Deok; Song, Jae Hoon; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    A preliminary analysis was performed based on the literatures and the patents for the neutron generators and targets for the lead slowing down time spectrometer (LSDTS) system. It was found that local neutron generator did not exhibit enough neutron intensity such as 1E+12 n/s, which is a minimum requirement for the LSDTS system to overcome curium backgrounds. However, a neutron generator implemented with an electron accelerator may provide a higher intensity around 1E+13 n/s and it is required to investigate further including a detail analysis. In addition to the neutron generator, a study on target was performed with the Monte Carlo simulation. In the study, an optimal design of target was suggested to provide a high neutron yield and a better thermal resistance. The suggested target consists several cylindrical plates with a certain cooling gap, which have increasing thickness and increasing radius.

  6. The alanine detector in BNCT dosimetry: dose response in thermal and epithermal neutron fields.

    Science.gov (United States)

    Schmitz, T; Bassler, N; Blaickner, M; Ziegner, M; Hsiao, M C; Liu, Y H; Koivunoro, H; Auterinen, I; Serén, T; Kotiluoto, P; Palmans, H; Sharpe, P; Langguth, P; Hampel, G

    2015-01-01

    The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model. The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. The alanine detector can be used without

  7. Chlorine detection in fly ash concrete using a portable neutron generator.

    Science.gov (United States)

    Naqvi, A A; Kalakada, Zameer; Al-Matouq, Faris A; Maslehuddin, M; Al-Amoudi, O S B

    2012-08-01

    The chlorine concentration in chloride-contaminated FA cement concrete specimens was measured using a portable neutron generator based prompt gamma-ray neutron activation (PGNAA) setup with the neutron generator and the gamma-ray detector placed side-by-side on one side of the concrete sample. The minimum detectable concentration of chlorine in FA cement concrete measured in the present study was comparable with previous results for larger accelerator based PGNAA setup. It shows the successful application of a portable neutron generator in concrete corrosion studies. Copyright © 2012 Elsevier Ltd. All rights reserved.

  8. Development of quasi-monochromatic p-7Li neutron generating system for 80-210 MeV

    International Nuclear Information System (INIS)

    Nakao, Noriaki; Shibata, Tokushi; Nakamura, Takashi; Uwamino, Yoshitomo; Nakanishi, Noriyoshi; Kurosawa, Tadahiro; Kim, Unju.

    1996-01-01

    Recently the requirements for the experimental data on the response characteristics of neutron detector and the cross section for neutron generation by charged particles have been increasing for shield designing. Here, a system for quasi-monochromatic neutron generation was developed in the facility of ring-cyclotron in Institute of Physical and Chemical Sciences. In this study, H 2 + accelerated to an energy range of 80-135 MeV/n and P + to 150-210 MeV was irradiated to E4 beam course and NE102A plastic scintillator was used for monitoring the neutron flux. The amount of neutrons generated was estimated from the radioactivity of 7 Be produced in 7 Li-target. The neutron spectres thus estimated as an energy range of 80-210 MeV were presented and the lower limit of these spectres was about 20 MeV. The peaks in the range of 150 and 210 MeV were comparatively wide because of the inferiority of energy resolving power at a higher energy level. (M.N.)

  9. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H p (10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. (authors)

  10. A neutron monitor for D-T neutron generator in the PGNAA-based online measurement system

    Science.gov (United States)

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

    2017-06-01

    A new type of neutron detector, which consists of polyethylene, an EJ200 plastic scintillator and fused silica, was proposed and optimized by the GEANT4 Monte Carlo simulation toolkit in our previous studies. The calculation method was also described for calculating the neutron flux in the preset condition. This paper reports the manufacturing of the prototype detector. Experiments are conducted to validate the feasibility of this detector. A D-T neutron generator and a 60Co gamma-ray source are used in the experiments. The designed detector and a He-3 proportional counter are simultaneously used to monitor the yield of the D-T neutron generator. A more universal calculation method is developed to enable the application of this detector to common conditions. The experimental results show that the performance of the designed detector is comparable to that of the He-3 proportional counter. The relative deviations between their normalized counts are less than 5%.

  11. A High Intensity Multi-Purpose D-D Neutron Generator for Nuclear Engineering Laboratories

    International Nuclear Information System (INIS)

    Ka-Ngo Leung; Jasmina L. Vujic; Edward C. Morse; Per F. Peterson

    2005-01-01

    This NEER project involves the design, construction and testing of a low-cost high intensity D-D neutron generator for teaching nuclear engineering students in a laboratory environment without radioisotopes or a nuclear reactor. The neutron generator was designed, fabricated and tested at Lawrence Berkeley National Laboratory (LBNL)

  12. Survey of neutron spectra generated by the fission of heavy nuclei induced by fast neutrons

    International Nuclear Information System (INIS)

    Lovchikova, G.N.; Trufanov, A.M.

    1997-01-01

    A review of neutron fission spectra measurements is presented. This review and the results of this analysis was performed with the participation of the authors. It is shown that there is a need for additional measurements of the energy and angular distributions of secondary neutrons in order to improve the understanding of the neutron emission mechanism in fission. (author). 21 refs, 6 figs

  13. Next generation neutron scattering at Neutron Science Center project in JAERI

    International Nuclear Information System (INIS)

    Yamada, Yasusada; Watanabe, Noboru; Niimura, Nobuo; Morii, Yukio; Katano, Susumu; Aizawa, Kazuya; Suzuki, Jun-ichi; Koizumi, Satoshi; Osakabe, Toyotaka.

    1997-01-01

    Japan Atomic Energy Research Institute (JAERI) has promoted neutron scattering researches by means of research reactors in Tokai Research Establishment, and proposes 'Neutron Science Research Center' to develop the future prospect of the Tokai Research Establishment. The scientific fields which will be expected to progress by the neutron scattering experiments carried out at the proposed facility in the Center are surveyed. (author)

  14. Results and plans on the development of a pulsed neutron generator

    International Nuclear Information System (INIS)

    Sztaricskai, T.; Vasvary, L.; Petoe, G.

    1976-01-01

    Using the vacuum system of an old van de Graaff machine a new pulsed neutron generator has been developed. The block diagram, the scheme of generators arrangement and the electrode system of the ion bunching parts are shown

  15. Feasibility study on medical isotope production using a compact neutron generator.

    Science.gov (United States)

    Leung, Ka-Ngo; Leung, James K; Melville, Graeme

    2018-07-01

    Compact neutron generators can provide high flux of neutrons with energies ranging from thermal (0.025 eV) to 14 MeV. Recent measurements demonstrated high neutron yields from the D- 7 Li fusion reaction at an interaction energy of 500 keV. Using the D- 7 Li reaction and applying new advancements in high flux neutron generator technology along with the commercial availability of high voltage DC power supplies enables the production of useful quantities of radioisotopes for medical applications. Using the known neutron reaction cross-sections, it has been estimated that hundreds-to-thousands MBq (or tens-to-hundreds mCi) of 99 Mo, 225 Ac, 64 Cu and 67 Cu can be obtained from a compact high flux neutron generator. Copyright © 2018 Elsevier Ltd. All rights reserved.

  16. Use of a high repetition rate neutron generator for in vivo body composition measurements via neutron inelastic scattering

    International Nuclear Information System (INIS)

    Kehayias, J.J.; Ellis, K.J.; Cohn, S.H.; Weinlein, J.H.

    1986-01-01

    A small D-T neutron generator with a high pulse rate is used for the in vivo measurement of body carbon, oxygen and hydrogen. The core of the neutron generator is a 13 cm-long Zetatron tube pulsed at a rate of 10 kHz delivering 10 3 to 10 4 neutrons per pulse. A target-current feedback system regulates the source of the accelerator to assure constant neutron output. Carbon is measured by detecting the 4.44 MeV γ-rays from inelastic scattering. The short half-life of the 4.44 MeV state of carbon requires detection of the γ-rays during the 10 μs neutron pulse. Generators with low pulsing rate were found inappropriate for carbon measurements because of their low duty-cycle (high neutron output during the pulse). In vivo measurements were performed with normal volunteers using a scanning bed facility for a dose less than 25 mrem. This technique offers medical as well as general bulk analysis applications. 8 refs., 5 figs

  17. Generation of neutron standing waves at total reflection of polarized neutrons

    International Nuclear Information System (INIS)

    Aksenov, V.L.; Nikitenko, Yu.V.; Kozhevnikov, S.V.; Radu, F.; Kruijs, R.; Rekveldt, M.Th.

    1999-01-01

    The regime of neutron standing waves at reflection of polarized thermal neutrons from the structure glass/Cu (1000 A Angstrom)/Ti (2000 A Angstrom)/Co (60 A Angstrom)/Ti (300 A Angstrom) in a magnetic field directed at an angle to the sample plane is realized. The intensity of neutrons with a particular spin projection on the external magnetic field direction appears to be a periodic function of the neutron wavelength and the glancing angle of the reflected beam. It is shown that the neutron standing wave regime can be a very sensitive method for the determination of changes in the spatial position of magnetic noncollinear layers. (author)

  18. 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...

  19. Study of response of 3He detectors to monoenergetic neutrons

    International Nuclear Information System (INIS)

    Abanades, A.; Andriamonje, S.; Arnould, H.; Barreau, G.; Bercion, M.; Casagrande, F.; Cennini, P.; Del Moral, R.; Gonzales, E.; Lacoste, V.; Pdemay, G.; Pravikoff, M.S.

    1997-01-01

    In the search of a hybrid system (the coupling of the particle accelerator to an under-critical reactor) for radioactive waste transmutation the TARC (Transmutation by Adiabatic Resonance Crossing) program has been developed. Due to experimental limitations, the time-energy relation at higher neutron energies, particularly, around 2 MeV, which is an important domain for TARC, cannot be applied. Consequently the responses of the 3 He ionization neutron detector developed for TARC experiment have been studied using a fast monoenergetic neutron source. The neutrons were produced by the interaction of the proton delivered by Van de Graaff accelerator of CENBG. The originality of the detector consists in its structure of three series of electric conductors which are mounted around the anode: a grid ensuring the detector proportionality, a cylindrical suit of alternating positive voltage and grounded wires aiming at eliminating the radial end effects, serving as veto and two cylinders serving as end plugs to eliminate the perpendicular end effects. Examples of anode spectra conditioned (in anticoincidence) by the mentioned vetoes are given. One can see the contribution of the elastic scattering from H and 3 He. By collimating the neutron beam through a borated polyethylene system it was possible to obtain a mapping of the detector allowing the study of its response as a function of the irradiated zones (anode and grid)

  20. A D-D/D-T fusion reaction based neutron generator system for liver tumor BNCT

    International Nuclear Information System (INIS)

    Koivunoro, H.; Lou, T.P.; Leung, K. N.; Reijonen, J.

    2003-01-01

    Boron-neutron capture therapy (BNCT) is an experimental radiation treatment modality used for highly malignant tumor treatments. Prior to irradiation with low energetic neutrons, a 10B compound is located selectively in the tumor cells. The effect of the treatment is based on the high LET radiation released in the 10 B(n,α) 7 Li reaction with thermal neutrons. BNCT has been used experimentally for brain tumor and melanoma treatments. Lately applications of other severe tumor type treatments have been introduced. Results have shown that liver tumors can also be treated by BNCT. At Lawrence Berkeley National Laboratory, various compact neutron generators based on D-D or D-T fusion reactions are being developed. The earlier theoretical studies of the D-D or D-T fusion reaction based neutron generators have shown that the optimal moderator and reflector configuration for brain tumor BNCT can be created. In this work, the applicability of 2.5 MeV neutrons for liver tumor BNCT application was studied. The optimal neutron energy for external liver treatments is not known. Neutron beams of different energies (1eV < E < 100 keV) were simulated and the dose distribution in the liver was calculated with the MCNP simulation code. In order to obtain the optimal neutron energy spectrum with the D-D neutrons, various moderator designs were performed using MCNP simulations. In this article the neutron spectrum and the optimized beam shaping assembly for liver tumor treatments is presented

  1. IEC-based neutron generator for security inspection system

    International Nuclear Information System (INIS)

    Miley, G.H.; Wu, L.; Kim, H.J.

    2005-01-01

    Use of a combined X-ray and neutron source for security inspections based on Inertial Electrostatic Confinement (IEC) fusion is discussed. Current inspection systems typically use X-ray techniques, but thermal neutron analysis (TNA) and fast neutron analysis (FNA), allow expanded detection of certain types of explosives. The integrated unit proposed here uses three separate IEC sources producing 14 and 2.45 MeV neutrons plus soft X-rays. This combination allows multiple detection methods with the composite signal analysis being done by a fuzzy logic system, significantly reducing false signals. (author)

  2. Response of neutron-irradiated RPV steels to thermal annealing

    International Nuclear Information System (INIS)

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels

  3. Neutron sensitivity of prompt-response self-powered neutron detectors and the interval rule

    International Nuclear Information System (INIS)

    Molina Avila, J.; Carmolopes, M.

    1989-01-01

    This paper is devoted to the calculation of thermal s th and epithermal s epi sensitivities of cobalt prompt-response Self-Powered Neutron Detectors (SPNDs). The thermal sensitivity was obtained for a Maxwellian neutron field, and the effect of scattering on the self-shielding correction was taken into consideration in the second-collision approximation. The dependence of s th on the emitter radius R was studied in a wide region of R (0.025 to 0.2 cm). The differential and global epithermal sensitivities were calculated using a simple expression for the first-collision neutron absorption probability. Finally, a criterion to evaluate the accuracy of the parameters of the model was established in the form of some Interval Rule which is very sensitive to the radial dependence of the flux perturbation correction and other parameters of the model in both the thermal and epithermal regions

  4. NEUTRON GENERATOR FACILITY AT SFU: GEANT4 DOSE RATE PREDICTION AND VERIFICATION.

    Science.gov (United States)

    Williams, J; Chester, A; Domingo, T; Rizwan, U; Starosta, K; Voss, P

    2016-11-01

    Detailed dose rate maps for a neutron generator facility at Simon Fraser University were produced via the GEANT4 Monte Carlo framework. Predicted neutron dose rates throughout the facility were compared with radiation survey measurements made during the facility commissioning process. When accounting for thermal neutrons, the prediction and measurement agree within a factor of 2 or better in most survey locations, and within 10 % inside the vault housing the neutron generator. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  5. Report of the advisory group meeting on optimal use of accelerator-based neutron generators

    International Nuclear Information System (INIS)

    1998-01-01

    During the past 20 to 25 years, the IAEA has provided a number of laboratories in the developing member states with neutron generators. These neutron generators were originally supplied for the primary purpose of neutron activation analysis. In order to promote the optimal use of these machines, a meeting was held in 1996, resulting in a technical document manual for the upgrading and troubleshooting of neutron generators. The present meeting is a follow-up to that earlier meeting. There are several reasons why some neutron generators are not fully utilized. These include lack of infrastructure, such as an appropriate shielded building and loss of adequately trained technical and academic personnel. Much of the equipment is old and lacking spare parts, and in a few cases there is a critical lack of locally available knowledge and experience in accelerator technology. The report contains recommendations for dealing with these obstacles

  6. Heat generation and temperature-rise in ordinary concrete due to capture of thermal neutrons

    International Nuclear Information System (INIS)

    Abdo, E.A.; Amin, E.

    1997-01-01

    The aim of this work is the evaluation of the heat generation and temperature-rise in local ordinary concrete as a biological shield due to capture of total thermal and reactor thermal neutrons. The total thermal neutron fluxes were measured and calculated. The channel number 2 of the ETRR-1 reactor was used in the measurements as a neutron source. Computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 was used in the calculations. The heat generation and temperature-rise in local ordinary concrete were evaluated and calculated. The results were displayed in curves to show the distribution of thermal neutron fluxes and heat generation as well as temperature-rise with the shield thickness. The results showed that, the heat generation as well as the temperature-rise have their maximum values in the first layers of the shield thickness. 4 figs., 12 refs

  7. Neutron flux response to regulating rod random vibrations

    International Nuclear Information System (INIS)

    Dach, K.; Nemec, J.; Pecinka, L.

    The relation is presented for the mean square value of the deflection of the rod for the n-th vibration shape on an arbitrary site. The relation may serve the obtaining of a variable which may be used both in a mechanical, i.e., stress analysis and in the determination of neutron flux fluctuations. It is demonstrated that the vibration frequency introduced in the reactor by the regulating rod has the same response in the neutron flux. This effect was used in the localization of an enormously vibrating regulating rod. (J.P.)

  8. Acoustic response of superheated droplet detectors to neutrons

    International Nuclear Information System (INIS)

    Gao Size; Zhang Guiying; Ni Bangfa; Zhao Changjun; Zhang Huanqiao; Guan Yongjing; Chen Zhe; Xiao Caijin; Liu Chao; Liu Cunxiong

    2012-01-01

    The search for dark matter (DM) is a hot field nowadays, a number of innovative techniques have emerged. The superheated droplet technique is relatively mature; however, it is recently revitalized in a number of frontier fields including the search for DM. In this work, the acoustic response of Superheated Droplet Detectors (SDDs) to neutrons was studied by using a 252 Cf neutron source, SDDs developed by the China Institute of Atomic Energy, a sound sensor, a sound card and a PC. Sound signals were filtered. The characteristics of FFT spectra, power spectra and time constants were used to determine the authenticity of the bubbles analyzed.

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

    Science.gov (United States)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

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

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

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    2016-01-01

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

  11. Neutron generator production mission in a national laboratory.

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Larry E.

    2007-08-01

    In the late 1980's the Department of Energy (DOE) faced a future budget shortfall. By the spring of 1991, the DOE had decided to manage this problem by closing three production plants and moving production capabilities to other existing DOE sites. As part of these closings, the mission assignment for fabrication of War Reserve (WR) neutron generators (NGs) was transferred from the Pinellas Plant (PP) in Florida to Sandia National Laboratories, New Mexico (SNL/NM). The DOE directive called for the last WR NG to be fabricated at the PP before the end of September 1994 and the first WR NG to be in bonded stores at SNL/NM by October 1999. Sandia National Laboratories successfully managed three significant changes to project scope and schedule and completed their portion of the Reconfiguration Project on time and within budget. The PP was closed in October 1995. War Reserve NGs produced at SNL/NM were in bonded stores by October 1999. The costs of the move were recovered in just less than five years of NG production at SNL/NM, and the annual savings today (in 1995 dollars) is $47 million.

  12. Descartes: a new generation system for neutronic calculations

    International Nuclear Information System (INIS)

    Calvin, Ch.

    2005-01-01

    Descartes is a common project between CEA, Framatome and EDF for the development of a new generation system for neutronic calculations. The main objectives which have leaded the design of the platform are the following: - flexible: from best-estimate calculations to industrial design; - open: easy coupling with other disciplines (thermo mechanics, thermal hydraulics); - enlarged scope: criticality, shielding, all types of reactors; - robust: well known behavior in its field of application; - safe: qualified and uncertainties assessment; and - User-friendly: user interface, databases; Descartes is based on the object oriented method using UML design and programmed in C++ and the Python interpreted script language. We will present in this paper the general architecture of the platform and the internal data model used which allows the definition of common exchange structures between solvers and the different modules which can be used either for lattice or core calculations. In a second time we will present a short description of the main solvers implemented within the Descartes platform. We will conclude with some first results of industrial PWR calculations. (author)

  13. Improv Chat: Second Response Generation for Chatbot

    OpenAIRE

    Wei, Furu

    2018-01-01

    Existing research on response generation for chatbot focuses on \\textbf{First Response Generation} which aims to teach the chatbot to say the first response (e.g. a sentence) appropriate to the conversation context (e.g. the user's query). In this paper, we introduce a new task \\textbf{Second Response Generation}, termed as Improv chat, which aims to teach the chatbot to say the second response after saying the first response with respect the conversation context, so as to lighten the burden ...

  14. X-Ray Measurements Of A Thermo Scientific P385 DD Neutron Generator

    International Nuclear Information System (INIS)

    Wharton, C. J.; Seabury, E. H.; Chichester, D. L.; Caffrey, A. J.; Simpson, J.; Lemchak, M.

    2011-01-01

    Idaho National Laboratory is experimenting with electrical neutron generators, as potential replacements for californium-252 radioisotopic neutron sources in its PINS prompt gamma-ray neutron activation analysis (PGNAA) system for the identification of military chemical warfare agents and explosives. In addition to neutron output, we have recently measured the x-ray output of the Thermo Scientific P385 deuterium-deuterium neutron generator. X rays are a normal byproduct from neutron generators, but depending on their intensity and energy, x rays can interfere with gamma rays from the object under test, increase gamma-spectrometer dead time, and reduce PGNAA system throughput. The P385 x-ray energy spectrum was measured with a high-purity germanium (HPGe) detector, and a broad peak is evident at about 70 keV. To identify the source of the x rays within the neutron generator assembly, it was scanned by collimated scintillation detectors along its long axis. At the strongest x-ray emission points, the generator also was rotated 60 deg. between measurements. The scans show the primary source of x-ray emission from the P385 neutron generator is an area 60 mm from the neutron production target, in the vicinity of the ion source. Rotation of the neutron generator did not significantly alter the x-ray count rate, and its x-ray emission appears to be axially symmetric. A thin lead shield, 3.2 mm (1/8 inch) thick, reduced the 70-keV generator x rays to negligible levels.

  15. Response function measurement of plastic scintillator for high energy neutrons

    International Nuclear Information System (INIS)

    Sanami, Toshiya; Ban, Syuichi; Takahashi, Kazutoshi; Takada, Masashi

    2003-01-01

    The response function and detection efficiency of 2''φ x 2''L plastic (PilotU) and NE213 liquid (2''NE213) scintillators, which were used for the measurement of secondary neutrons from high energy electron induced reactions, were measured at Heavy Ion Medical Accelerator in Chiba (HIMAC). High energy neutrons were produced via 400 MeV/n C beam bombardment on a thick graphite target. The detectors were placed at 15 deg with respect to C beam axis, 5 m away from the target. As standard, a 5''φ x 5''L NE213 liquid scintillator (5''NE213) was also placed at same position. Neutron energy was determined by the time-of-flight method with the beam pickup scintillator in front of the target. In front of the detectors, veto scintillators were placed to remove charged particle events. All detector signals were corrected with list mode event by event. We deduce neutron spectrum for each detectors. The efficiency curves for pilotU and 2''NE213 were determined on the bases of 5 N E213 neutron spectrum and its efficiency calculated by CECIL code. (author)

  16. Calculation of the importance-weighted neutron generation time using MCNIC method

    International Nuclear Information System (INIS)

    Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

    2008-01-01

    In advanced nuclear power systems, such as ADS, the need for reliable kinetics parameters is of considerable importance because of the lower value for β eff due to the large amount of transuranic elements loaded in the core of those systems. All reactor kinetic parameters are weighted quantities. In other words each neutron with a given position and energy is weighted with its importance. Neutron generation time as an important kinetic parameter, in all nuclear power systems has a significant role in the analysis of fast transients. The difference between non-weighted neutron generation time; Λ; standard in most Monte Carlo codes; and the weighted one Λ + can be quite significant depending on the type of the system. In previous work, based on the physical concept of neutron importance, a new method; MCNIC; using the MCNP code has been introduced for the calculation of neutron importance in fissionable assemblies for all criticality states. In the present work the applicability of MCNIC method has been extended for the calculation of the importance-weighted neutron generation time. The influence of reflector thickness on importance-weighted neutron generation time has been investigated by the development of an auxiliary code, IWLA, for a hypothetic assembly. The results of these calculations were compared with the non-weighted neutron generation times calculated using the Monte Carlo code MCNP. The difference between the importance-weighted and non-weighted quantity is more significant in a reflected system and increases with reflector thickness

  17. The applicability of detailed process for neutron resonance absorption to neutronics analyses in LWR next generation fuels to extend burnup

    International Nuclear Information System (INIS)

    Kameyama, Takanori; Nauchi, Yasushi

    2004-01-01

    Neutronics analyses with detail processing for neutron resonance absorption in LWR next generation UOX and MOX fuels to extend burnup were performed based on the neutronic transport and burnup calculation. In the detailed processing, ultra-fine energy nuclear library and collision probabilities between neutron and U, Pu nuclides (actinide nuclides) are utilized for two-dimension geometry. In the usual simple processing (narrow resonance approximation), shielding factors and compensation equations for neutron resonance absorption are utilized. The results with detailed and simple processing were compared to clarify where the detailed processing is needed. The two processing caused difference of neutron multiplication factor by 0.5% at the beginning of irradiation, while the difference became smaller as burnup increased and was not significant at high burnup. The nuclide compositions of the fuel rods for main actinide nuclides were little different besides Cm isotopes by the processing, since the neutron absorption rate of 244 Cm became different. The detail processing is needed to evaluate the neutron emission rate in spent fuels. In the fuel assemblies, the distributions of rod power rates were not different within 0.5%, and the peak rates of fuel rod were almost the same by the two processing at the beginning of irradiation when the peak rate is the largest during the irradiation. The simple processing is also satisfied for safety evaluation based on the peak rate of rod power. The difference of local power densities in fuel pellets became larger as burnup increased, since the neutron absorption rate of 238 U in the peripheral region of pellets were significantly different by the two processing. The detail processing is needed to evaluate the fuel behavior at high burnup. (author)

  18. EXPERIMENTAL ANALYSES OF SPALLATION NEUTRONS GENERATED BY 100 MEV PROTONS AT THE KYOTO UNIVERSITY CRITICAL ASSEMBLY

    Directory of Open Access Journals (Sweden)

    CHEOL HO PYEON

    2013-02-01

    Full Text Available Neutron spectrum analyses of spallation neutrons are conducted in the accelerator-driven system (ADS facility at the Kyoto University Critical Assembly (KUCA. High-energy protons (100 MeV obtained from the fixed field alternating gradient accelerator are injected onto a tungsten target, whereby the spallation neutrons are generated. For neutronic characteristics of spallation neutrons, the reaction rates and the continuous energy distribution of spallation neutrons are measured by the foil activation method and by an organic liquid scintillator, respectively. Numerical calculations are executed by MCNPX with JENDL/HE-2007 and ENDF/B-VI libraries to evaluate the reaction rates of activation foils (bismuth and indium set at the target and the continuous energy distribution of spallation neutrons set in front of the target. For the reaction rates by the foil activation method, the C/E values between the experiments and the calculations are found around a relative difference of 10%, except for some reactions. For continuous energy distribution by the organic liquid scintillator, the spallation neutrons are observed up to 45 MeV. From these results, the neutron spectrum information on the spallation neutrons generated at the target are attained successfully in injecting 100 MeV protons onto the tungsten target.

  19. A D-D neutron generator using a titanium drive-in target

    International Nuclear Information System (INIS)

    Kim, I.J.; Jung, N.S.; Jung, H.D.; Hwang, Y.S.; Choi, H.D.

    2008-01-01

    A D-D neutron generator was developed with an intensity of 10 8 n/s. A helicon plasma ion source was used to produce a large current deuteron beam, and neutrons were generated by irradiating the deuteron beam on a titanium drive-in target made of commercial pure titanium. The neutron generator was test-run for several hundred hours, and the performances were investigated. The available range of the deuteron beam current was 0.8-8 mA and the beam could be accelerated up to 97.5 keV. The maximum neutron generation rate in the test-runs was 1.9 x 10 8 n/s, which was achieved by irradiating a 7.6 mA deuteron beam at 94.0 keV on a 0.5 mm-thick target. The operation of the neutron generator was fairly stable, such that the neutron generation rate was not altered by high voltage breakdowns during the test-runs. Neutron generation efficiency was rated as low as 10% when compared to an ideal case of irradiating a 100% monatomic deuteron beam on a perfect TiD 2 target. Factors causing the low efficiency were suggested and discussed

  20. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

    1994-01-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

  1. COMBINE/PC - a portable neutron spectrum and cross-section generation program

    International Nuclear Information System (INIS)

    Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

    1990-01-01

    Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure

  2. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)

    2013-09-15

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.

  3. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    International Nuclear Information System (INIS)

    Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.

    2013-01-01

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

  4. DETERMINATION OF LIMIT DETECTION OF THE ELEMENTS N, P, K, Si, Al, Fe, Cu, Cd, WITH FAST NEUTRON ACTIVATION USING NEUTRON GENERATOR

    OpenAIRE

    Sunardi, Sunardi; Muryono, Muryono

    2010-01-01

    Determination of limit detection of the elements N, P, K, Si, Al, Fe, Cu, Cd, with fast neutron activation using neutron generator has been done.  Samples prepared from SRM 2704, N, P, K elements from MERCK, Cu, Cd, Al from activation foil made in San Carlos, weighted and packed for certain weight then iradiated during 30 minutes with 14 MeV fast neutron using the neutron generator and then counted with gamma spectrometry (accuspec).  At this research condition of neutron generator was set at...

  5. Neutron generators and their uses in research and applied fields. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    Asfour, F I [Division of Basic Nuclear Sciences, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    The development of the low voltage neutron generators (NGS) has contributed considerably to the scope of nuclear research and the economical application of nuclear methods. Such simple instruments are used to produce 14 MeV and 3 MeV neutrons via the 3{sup H}(d,n)4{sup H}e and 2{sup H}(d,n)3{sup H}e reactions,respectively. The neutrons are very widely used and are inexpensive, easy to install and operate, therefore, in addition to nuclear physicists, there are a number of groups of scientists who use low voltage accelerators as tools for pure and applied research, service and education. The aim of this work is to review shortly those problems and methods of science and technology where the neutrons produced in the D-T and D-D reactions play the main role. A wide range of experiments with the detection of neutrons and charged particles is available including the study of shielding and the generator technology itself. N.G. are recently widely used for the determination of neutron data needed for fast reactor and thermonuclear devices. The principles and techniques of the possible uses of neutron generators in technology and research are summarized. The review is devoted to:- Give a short review of the most important operational characteristics of the neutron generators and the necessary instruments needed for application. Outline the main applications of the neutron generators in neutron activation and prompt radiation analysis in various fields(metallurgy, chemistry, biology, meteoritic and lunar studies, geology and mining, etc...) fast neutron therapy, and radiation effects. 2 figs.

  6. The electric dipole response of neutron rich tin isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Horvat, Andrea; Aumann, Thomas; Rossi, Dominic; Schindler, Fabia [Institut fuer Kernphysik, TU Darmstadt (Germany); Boretzky, Konstanze [GSI Helmholzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Johansen, Jacob [Aarhus University (Denmark); Schrock, Philipp [The University of Tokyo (Japan); Collaboration: R3B-Collaboration

    2016-07-01

    Studies of the dipole response in medium heavy and heavy neutron rich nuclei reveal valuable information about the isospin dependence of the nuclear equation of state. Therefore an experimental campaign investigating both the electric dipole response via Coulomb excitation and neutron removal along the tin isotope chain ({sup 124-134}Sn) has been carried out at the R3B (Reactions with Relativistic Radioactive Beams) setup at GSI (Helmholtzzentrum fuer Schwerionenforschung) for which the analysis is ongoing. The E1 response was induced via relativistic Coulomb scattering by a lead target in inverse kinematics, and calls for a kinematically complete determination of all reaction products in order to reconstruct the excitation energy by means of the invariant mass method. The goal is to obtain the Coulomb excitation cross section up to the adiabatic cut-off energy, covering the giant dipole resonance (GDR) range.

  7. Isotope identification as a part of the decommissioning of San Diego State University's Texas Nuclear neutron generator

    International Nuclear Information System (INIS)

    Taylor, D.

    1997-07-01

    present in this type of unit. All activation products of neutron generator components, as a result of the exposure to 14 MeV neutrons are short lived, and would have decayed to below detection levels in the period of time since the unit was last operated, a period of approximately six years. This project consists of identifying the nuclide responsible for the apparent radiation energy. Additionally, as a service to San Diego State's Radiation Safety Office, an estimate of the amount of activity in the unit will be made, as required for future disposal purposes as well as packaging the unit for shipment according to current state and federal regulations

  8. NGI-9 pulsed neutron generator with a fluence to 1010 n/s

    International Nuclear Information System (INIS)

    Allakhverdov, A.Sh.; Ogarkin, V.I.; Silicheva, G.P.; Timofeev, Yu.I.

    1975-01-01

    A neutron pulse generator with 14 MeV energy used for the activation analysis, is described. Its functional diagram and the technical characteristics are presented. The studies of the generator that resulted in determination of the effect of the accelerating voltage amplitude, the delay in the ion source firing with respect to the pulse of the accelerating voltage, the amount of operating ion sources and the energy imparted to them on the neutron flux magnitude are conducted. It is confirmed by the studies that the neutron generator operating in the nominal regime makes it possible to obtain a neutron flux of 5x10 9 -10 10 neutr./s. The dependence of the neutron flux variation on the frequency of pulse sequence for various ion sources is shown

  9. Instrumentation system for pulsed neutron generator. Pt. 1. Electronic control and data acquisition

    International Nuclear Information System (INIS)

    Burda, J.; Igielski, A.; Janik, W.; Kosik, M.; Kurowski, A.; Zaleski, T.

    1997-01-01

    The paper presents an electronic instrumentation system which is successfully applied for pulsed neutron generator and measurements. In the paper there are described in details all modernized parts of the system as well as new designed and applied ones. The set of diagrams is enclosed. An important part of the system has been designed and built in the Neutron Transport Physics Laboratory. (author)

  10. Generation of neutron cross sections library for the Thermos code of the Fuel management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Viais J, J.

    1990-10-01

    There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)

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

    Science.gov (United States)

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

  12. Neutron production and ion beam generation in plasma focus devices

    International Nuclear Information System (INIS)

    Steinmetz, K.

    1980-01-01

    Concerning the physical processes leading to neutron emission, a clearer situation has been achieved compared to the state at the start of this work. The general discussion will realize that the whole experimental data cannot be described consistently by the predictions of either the beam-target model or the quasi-thermonuclear fusion model, although many questions about the neutron production properties have been solved. In particular the neutron fluence anisotropy is found to be a property basically related to the existence of fast ions escaping axially out of the pinch region. The requirements to explain broad radial neutron energy spectra, long emission times, and energetic but not spatial emission anisotropies suggest a kind of particle trapping in the main source region. (orig./HT)

  13. Production of a pulseable fission-like neutron flux using a monoenergetic 14 MeV neutron generator and a depleted uranium reflector

    Science.gov (United States)

    Koltick, D.; McConchie, S.; Sword, E.

    2008-04-01

    The design and performance of a pulseable neutron source utilizing a D-T neutron generator and a depleted uranium reflector are presented. Approximately half the generator's 14 MeV neutron flux is used to produce a fission-like neutron spectrum similar to 252Cf. For every 14 MeV neutron entering the reflector, more than one fission-like neutron is reflected back across the surface of the reflector. Because delayed neutron production is more than two orders of magnitude below the prompt neutron production, the source takes full advantage of the generator's pulsed mode capability. Applications include all elemental characterization systems using neutron-induced gamma-ray spectroscopy. The source simultaneously emits 14 MeV neutrons optimal to excite fast neutron-induced gamma-ray signals, such as from carbon and oxygen, and fission-like neutrons optimal to induce neutron capture gamma-ray signals, such as from hydrogen, nitrogen, and chlorine. Experiments were performed, which compare well to Monte Carlo simulations, showing that the uranium reflector enhances capture signals by up to a factor of 15 compared to the absence of a reflector.

  14. Response of LR-115 type II and CR-39 plastic track detectors to Am-Be and 14.1-MeV neutrons

    International Nuclear Information System (INIS)

    Bradley, D.A.; Chong, C.S.; Saat, Ahmat; Sidik, A.G.; Ghose, A.M.

    1987-01-01

    The fast-neutron response of the plastic LR-115 type II and CR-39 track detectors have been compared, using a 14.1-MeV neutron generator and a radionuclide Am-Be neutron source (effective primary neutron energy 4.5-MeV). The distribution of track diameters for a range of etching times has been evaluated, taking into account track registration efficiency and the relevant fast neutron scattering cross-sections. The efficiency of etched-track formation in LR-115 type II due to neutron irradiation is approximately double that in CR-39. The 14.1-MeV neutrons also tend to produce tracks in both materials with somewhat greater efficiency than do the lower energy neutrons from the radionuclide source, for a given etching time. (author)

  15. Simulation and preliminary experimental results for an active neutron counter using a neutron generator for a fissile material accounting

    International Nuclear Information System (INIS)

    Ahn, Seong-Kyu; Lee, Tae-Hoon; Shin, Hee-Sung; Kim, Ho-Dong

    2009-01-01

    An active neutron coincidence counter using a neutron generator as an interrogation source has been suggested. Because of the high energy of the interrogation neutron source, 2.5 MeV, the induced fission rate is strongly affected by the moderator design. MCNPX simulation has been performed to evaluate the performance achieved with these moderators. The side- and bottom-moderator are significantly important to thermalize neutrons to induce fission. Based on the simulation results, the moderators are designed to be adapted to the experimental system. Their preliminary performance has been tested by using natural uranium oxide powder samples. For a sample of up to 3.5 kg, which contains 21.7 g of 235 U, 2.64 cps/g- 235 U coincidence events have been measured. Mean background error was 9.57 cps and the resultant coincidence error was 13.8 cps. The experimental result shows the current status of an active counting using a neutron generator which still has some challenges to overcome. However, the controllability of an interrogation source makes this system more applicable for a variety of combinations with other non-destructive methods like a passive coincidence counting especially under a harsh environment such as a hot cell. More precise experimental setup and tests with higher enriched samples will be followed to develop a system to apply it to an active measurement for the safeguards of a spent fuel treatment process.

  16. Determination for energy response and directionality of neutron survey meters

    International Nuclear Information System (INIS)

    Chen Changmao; Liu Jinhua; Xie Jianlun; Su Jingling

    1992-01-01

    The energy response and directionality of neutron survey meter type MK7 and 2202D are determined. The reactor thermal column beam, reactor filtered beams (6 eV, 24.4 keV and 144 keV), 226 Ra-Be, 241 Am-Be, 252 Cf and its moderated sources are used for the measurement. The results shows: the survey meters are influenced obviously by the direction; the response of middle-energy region is large, the energy response of 2202D is better than MK7

  17. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Directory of Open Access Journals (Sweden)

    Hegazy Aya Hamdy

    2018-01-01

    Full Text Available Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1 shielding-collimator material, (2 Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3 thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  18. Development of High Intensity D-T fusion NEutron Generator (HINEG)

    Science.gov (United States)

    Wu, Yican; Liu, Chao; Song, Gang; Wang, Yongfeng; Li, Taosheng; Jiang, Jieqiong; Song, Yong; Ji, Xiang

    2017-09-01

    A high intensity D-T fusion neutron generator (HINEG) is keenly needed for the research and development (R&D) of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014 1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

  19. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Science.gov (United States)

    Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.

    2018-04-01

    Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  20. Development of High Intensity D-T fusion NEutron Generator (HINEG

    Directory of Open Access Journals (Sweden)

    Wu Yican

    2017-01-01

    Full Text Available A high intensity D-T fusion neutron generator (HINEG is keenly needed for the research and development (R&D of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014~1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

  1. Associated-particle sealed-tube neutron generators and hodoscopes for NDA applications

    International Nuclear Information System (INIS)

    Rhodes, E.; Peters, C.W.

    1991-01-01

    With radioisotope sources, gamma-ray transmission hodoscopes can inspect canisters and railcars to monitor rocket motors, can detect nuclear warheads by their characteristic strong gamma-ray absorption, or can count nuclear warheads inside a missile by low-resolution tomography. Intrinsic gamma-ray radiation from warheads can also be detected in a passive mode. Neutron hodoscopes can use neutron transmission, intrinsic neutron emission, or reactions stimulated by a neutron source, in treaty verification roles. Gamma-ray and neutron hodoscopes can be combined with a recently developed neutron diagnostic probe system, based on a unique associated-particle sealed-tube neutron generator (APSTNG) that interrogates the object of interest with a low-intensity beam of 14-MeV neutrons, and that uses flight-time to electronically collimate transmitted neutrons and to tomographically image nuclides identified by reaction gamma-rays. Gamma-ray spectra of resulting neutron reactions identify nuclides associated with all major chemicals in chemical warfare agents, explosives, and drugs, as well as many pollutants and fissile and fertile special nuclear material. 5 refs., 12 figs

  2. 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)

  3. Symposium on CIAE 600 kV ns pulse neutron generator

    International Nuclear Information System (INIS)

    Shen Guanren

    2001-01-01

    CIAE 600 kV ns Pulse Neutron Generator was built by China National Nuclear Corporation, which is an important facility mainly used for experimental researches of nuclear reactions induced by 14 MeV neutrons, experimental measurements of energy spectra of secondary neutrons and charged particles and macro-checking experiments of evaluated neutron database and dosimetry researches of neutrons and γ rays. It is the first home made one, but the fourth similar facility in the world. Six articles are included in this symposium. The articles details the general structure, radio frequency ion source, high current beam ns pulsed system, etc. The main technical problems resolved during development are discussed. And attentions and experiences in the generator adjustments are introduced

  4. A study on measurement of neutrons generated in radiation therapy – Measurement of neurons in CR-39 detection method

    International Nuclear Information System (INIS)

    Park, Cheol-Soo; Cho, Jae-Hwan; Lee, Hae-Kag; Lee, Sun-Yeob; Jang, Hyon-Chol; Dong, Kyung-Rae; Chung, Woon-Kwan; Jin, Lee; Moon, Deog-Hwan; Lee, Kwang-Sung; Yang, Nam-Oh; Cho, Moo-Seong

    2013-01-01

    Highlights: ► To measure the neutrons generated in a linear accelerator. ► Both fast neutrons and thermal neutrons produced an increase in the dose of neutrons generated with increasing irradiation dose. ► The generation of neutrons increased when a wedge filter was used. ► When the SRS cone that required a high dose was used, more neutrons were detected. -- Abstract: The CR-39 [diethylene glycol bis-(allylcarbonate)] neuron detection method was used to measure the dose of neutrons generated in X-ray (photon) therapy conducted in a linear accelerator, and to use high-energy photons as part of the clinical applications to examine the problems associated with the dose for patients caused by the generation of neutrons from high-energy photons used for cancer therapy. According to the experimental results, 0.35 mSv, 0.65 mSv 1.82 mSv of fast neutrons on average were generated from 1 Gy, 2 Gy and 5 Gy of photon irradiation, respectively, whereas 0.26 mSv, 0.56 mSv and 1.23 mSv of thermal neutrons were generated. Both fast neutrons and thermal neutrons produced an increase in the dose of neutrons generated with increasing irradiation dose. With in regard to the dose generated within and around the irradiation area of the photon rays, it was confirmed that more neutrons were generated within the irradiation area. A wedge filer was used to measure the generation of neutrons. According to the measurement results, the generation of neutrons increased when a wedge filter was used. When the SRS cone that required a high dose was used, more neutrons were detected than those in the previous experiment. When fast neutrons were used, 2.85 mSv neutrons on average were generated from 5 Gy of photon irradiation. When thermal neutrons were used, 1.37 mSv neutrons on average were generated from 5 Gy of photon irradiation. Overall, approximately 1.6 times and 1.12 times more fast and thermal neutrons, respectively, were generated than in the case of a general treatment with 5 Gy

  5. The measurement of the fluorescence response of plastic scintillator ST1422 to neutron

    International Nuclear Information System (INIS)

    Luo Xiaobing; Xia Yijun; Yang Zhihua; Zhang Chuanfei; Peng Taiping; Li Rurong; Zhang Jianhua

    2004-01-01

    By using the T(p,n) and D(d,n) reactions as neutron resources, the fluorescence response of ST1422 plastic scintillators (40 mm in diameter and 5 mm in thickness) was determined in he neutron energy range of 0.655-5 MeV. The experimental results show a nonlinear response of fluorescence with neutron energy. (authors)

  6. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    International Nuclear Information System (INIS)

    Li Gang; Zhang Zhongshuai; Chi Qian; Liu Linmao

    2012-01-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 × 10 8 n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 μs. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  7. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    Energy Technology Data Exchange (ETDEWEB)

    Li Gang; Zhang Zhongshuai [Northeast Normal University, Changchun 130024 (China); Chi Qian [Guang Hua College of Chang Chun University, Changchun 130117 (China); Liu Linmao, E-mail: ll888@nenu.edu.cn [Northeast Normal University, Changchun 130024 (China)

    2012-11-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 Multiplication-Sign 10{sup 8} n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 {mu}s. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  8. Radioactive source recovery program responses to neutron source emergencies

    International Nuclear Information System (INIS)

    Dinehart, S.M.; Hatler, V.A.; Gray, D.W.; Guillen, A.D.

    1997-01-01

    Recovery of neutron sources containing Pu 239 and Be is currently taking place at Los Alamos National Laboratory. The program was initiated in 1979 by the Department of Energy (DOE) to dismantle and recover sources owned primarily by universities and the Department of Defense. Since the inception of this program, Los Alamos has dismantled and recovered more than 1000 sources. The dismantlement and recovery process involves the removal of source cladding and the chemical separation of the source materials to eliminate neutron emissions. While this program continues for the disposal of 239 Pu/Be sources, there is currently no avenue for the disposition of any sources other than those containing Pu 239 . Increasingly, there have been demands from agencies both inside and outside the Federal Government and from the public to dispose of unwanted sources containing 238 Pu/Be and 241 Am/Be. DOE is attempting to establish a formal program to recover these sources and is working closely with the Nuclear Regulatory Commission (NRC) on a proposed Memorandum of Understanding to formalize an Acceptance Program. In the absence of a formal program to handle 238 Pu/Be and 241 Am/Be neutron sources, Los Alamos has responded to several emergency requests to receive and recover sources that have been determined to be a threat to public health and safety. This presentation will: (1) review the established 239 Pu neutron source recovery program at Los Alamos, (2) detail plans for a more extensive neutron source disposal program, and (3) focus on recent emergency responses

  9. Responses of commercially available neutron electronic personal dosemeters in neutron fields simulating workplaces at MOX fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    The authors investigated the performance of three commercially available electronic personal dosemeters (EPDs) in evaluating neutron dose equivalents and discussed their suitability to work environments in MOX fuel fabrication facilities. The EPDs selected for this study were NRY21 (Fuji Electric Systems), PDM-313 (Aloka) and DMC 2000 GN (MGP Instruments). All tests were conducted in moderated 252 Cf neutron fields with neutron spectral and dosimetric characteristics similar to those found in MOX fuel facilities. The test results revealed trends and the magnitude of response variations in relation to neutron spectral changes expected in work environments.

  10. 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.

  11. The applied research program of the High Flux Neutron Generator at the National Nuclear Center, Havana

    International Nuclear Information System (INIS)

    Perez, G.; Martin, G.; Ceballos, C.; Padron, I.; Shtejer, K.; Perez, N.; Guibert, R.; Ledo, L.M.; Cruz Inclan, Carlos

    2001-01-01

    The Havana High Flux Neutron Generator facility is an intense neutron source based on a 20 mA duoplasmatron ion source and a 250 kV high voltage power supply. It has been installed in the Neutron Generator Laboratory at the Center of Applied Technologies and Nuclear Research in 1997. This paper deal outlined the future applied program to be carried out in this facility in the next years. The Applied Research Program consists on install two nuclear analytic techniques: the PELAN technique which uses the neutron generator in the pulse mode and the Low Energy PIXE technique which uses the same facility as a low energy proton accelerator for PIXE analysis. (author)

  12. Nanotubes based neutron generator for calibration of neutrino and dark matter detectors

    Science.gov (United States)

    Chepurnov, A. S.; Ionidi, V. Y.; Kirsanov, M. A.; Kitsyuk, E. P.; Klenin, A. A.; Kubankin, A. S.; Oleinik, A. N.; Pavlov, A. A.; Shchagin, A. V.

    2017-12-01

    The compact 2.45 MeV fast neutron generator with a reduced supply voltage for calibration of low-background neutrino and dark matter detectors was tested. The generator is based on an array of carbon nanotubes. Neutron generation is carried out by applying a high voltage in the range of +10 to + 25 kV to a nanotube array, which cause an ionization of deuterium molecules with the following acceleration of ions in the direction of the grounded target covered by a deuterated polyethylene film. The d(d,n)3He nuclear reaction happens as the result of ions collisions with the target. The dependences of the neutron yield as functions of the applied voltage were obtained for two different types of carbon nanotubes array. It is shown that the type of nanotubes array does not influence significantly on the neutron yield.

  13. Monte Carlo simulations of a D-T neutron generator shielding for landmine detection

    International Nuclear Information System (INIS)

    Reda, A.M.

    2011-01-01

    Shielding for a D-T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other 'threat' materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. - Highlights: → A landmine detection system based on neutron fast/slow analysis has been designed. → Shielding for a D-T sealed neutron generator tube has been designed using Monte Carlo radiation transport code. → Detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. → The signal-to-background ratio optimized at one position for all depths.

  14. DOSIMETRIC response of a REM-500 in low energy neutron fields typical of nuclear power plants.

    Science.gov (United States)

    Aslam; Matysiak, W; Atanackovic, J; Waker, A J

    2012-06-01

    ~0.3 at 51 keV neutron energy. The counter response based on ICRP 60 was comparable to an ideal response of 1.0 above 600 keV, which dropped to ~0.8 at 159 keV and ~0.4 at 51 keV neutron energy. The decline in counter quality factor response based on ICRP 60 was found to be much steeper than that when using the instrument’s built-in function for quality factor.The REM-500 measures a dose equivalent at 727 keV,which is 60% of the ambient dose equivalent, 40% at 159 keV,and 15% at 51 keV. Two algorithms have been developed, one for real time measurement and another to be used post measurement,and their efficacy is demonstrated in determining the quality factor and the ambient dose equivalent in low energy neutron fields, which are typical for the workplace environment in CANDU® nuclear power generating stations.

  15. DETERMINATION OF LIMIT DETECTION OF THE ELEMENTS N, P, K, Si, Al, Fe, Cu, Cd, WITH FAST NEUTRON ACTIVATION USING NEUTRON GENERATOR

    Directory of Open Access Journals (Sweden)

    Sunardi Sunardi

    2010-06-01

    Full Text Available Determination of limit detection of the elements N, P, K, Si, Al, Fe, Cu, Cd, with fast neutron activation using neutron generator has been done.  Samples prepared from SRM 2704, N, P, K elements from MERCK, Cu, Cd, Al from activation foil made in San Carlos, weighted and packed for certain weight then iradiated during 30 minutes with 14 MeV fast neutron using the neutron generator and then counted with gamma spectrometry (accuspec.  At this research condition of neutron generator was set at current 1 mA that produced neutron flux about 5,47.107 n/cm2.s and  experimental result shown that the limit detection for the elements N, P, K, Si, Al, Fe, Cu, Cd are  2,44 ppm, 1,88 ppm, 2,15 ppm, 1,44 ppm, 1,26 ppm, 1,35 ppm, 1,05 ppm, 2,99 ppm, respectively.  The data  indicate that the limit detection or sensitivity of appliance of neutron generator to analyze the element is very good, which is feasible to get accreditation AANC laboratory using neutron generator.   Keywords: limit detection, AANC, neutron generator

  16. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    International Nuclear Information System (INIS)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C.

    2017-01-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S N ) formulation and the response matrix method was applied to solve the forward and the adjoint S N problems. Numerical results are given to verify the present. (author)

  17. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: lrcmoraes@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: ricardob@iprj.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pós-Graduação em Modelagem Computacional

    2017-07-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S{sub N}) formulation and the response matrix method was applied to solve the forward and the adjoint S{sub N} problems. Numerical results are given to verify the present. (author)

  18. Loss of the associated α-particles in the tagged neutron generators

    Energy Technology Data Exchange (ETDEWEB)

    Sudac, D.; Nad, K.; Obhodas, J. [Institute Ruder Boskovic, P.O. Box 180, 10002 Zagreb (Croatia); Bystritsky, V.M. [Joint Institute for Nuclear Research, Moscow region, Dubna 141 980 (Russian Federation); Valkovic, V., E-mail: valkovic@irb.hr [Kvinticka 62, 10000 Zagreb (Croatia)

    2015-09-01

    The reported loss of α-particles in the 14 MeV tagged neutron generators has been investigated using two neutron generators equipped with α-particle counters and two neutron detectors. One neutron detector was put right in the middle of the tagged neutron cone and another one was put outside the cone. By measuring the difference between double (neutron–neutron) and triple (α-neutron-neutron) coincidences it is possible to deduce the α-particle loss since the number of triple coincidences should be equal to the number of double coincidences. In all measurements performed a deficit of triple with respect to double coincidences has been observed. This deficit was smallest for the threshold of α-particle Constant Fraction Discriminator (αCFD) being 0 and maximum allowed voltage of α-particle detector being −1.7 kV. The smallest measured deficit value was equal to 13±1%. From the observed results it was concluded that the deficit was due to a number of non-detected α-particles that loose sufficient quantity of energy while traveling to the detector because of collisions with particles present in the neutron tube and/or in the tritium target. These α-particles will not be detected as they fall under the threshold of αCFD discriminator. Magnetic fields present in the system worsen the situation since they are forcing α-particles to travel larger distances because of toroidal movement and undergoing additional collisions. Tagged neutron technique has many kind of applications and it is particularly important for high accuracy nuclear cross-sections measurements when α-particles losses must be carefully assessed.

  19. Generation of broad-group neutron/photon cross-section libraries for shielding applications

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Roussin, R.W.; Fu, C.Y.; White, J.E.

    1989-01-01

    The generation and use of multigroup cross-section libraries with broad energy group structures is primarily for the economy of computer resources. Also, the establishment of reference broad-group libraries is desirable in order to avoid duplication of effort, both in terms of the data generation and verification, and to assure a common data base for all participants in a specific project. Uncertainties are inevitably introduced into the broad-group cross sections due to approximations in the grouping procedure. The dominant uncertainty is generally with regard to the energy weighting function used to average the pointwise or fine-group data within a single broad group. Intelligent choice of the weighting functions can reduce such uncertainties. Also, judicious selection of the energy group structure can help to reduce the sensitivity of the computed responses to the weighting function, at least for a selected set of problems. Two new multigroup cross section libraries have been recently generated from ENDF/B-V data for two specific shielding applications. The first library was prepared for use in sodium-cooled reactor systems and is available in both broad-group structures. The second library, just recently completed, was prepared for use in air-over-ground environments and is available in a broad-group (46-neutron, 23-photon) energy structure. The selection of the specific group structures and weighting functions was an important part of the generation of both libraries

  20. New perspectives from new generations of neutron sources

    Science.gov (United States)

    Mezei, Ferenc

    2007-09-01

    Since the early 1950s the vital multidisciplinary progress in understanding condensed matter is, in a substantial fraction, based on results of neutron scattering experiments. Neutron scattering is an inherently intensity limited method and after 50 years of considerable advance—primarily achieved by improving the scattering instruments—the maturation of the technique of pulsed spallation sources now opens up the way to provide more neutrons with improved cost and energy efficiency. A quantitative analysis of the figure-of-merit of the specialized instruments for pulsed source operation shows that up to 2 orders of magnitude intensity gains can be achieved in the next decade, with the advent of high power spallation sources. The first stations on this road, the MW class short pulse spallation sources SNS in the USA (under commissioning), and J-PARC in Japan (under construction) will be followed by the 5 MW long pulse European Spallation Source (ESS). Further progress, that can be envisaged on the longer term, could amount to as much as another factor of 10 improvement. To cite this article: F. Mezei, C. R. Physique 8 (2007).

  1. New perspectives from new generations of neutron sources

    International Nuclear Information System (INIS)

    Mezei, F.

    2007-01-01

    Since the early fifties the vital multidisciplinary progress in understanding condensed matter is, in a substantial fraction, based on results of neutron scattering experiments. Neutron scattering is an inherently intensity limited method and after 50 years of considerable advance - primarily achieved by improving the scattering instruments - the maturation of the technique of pulsed spallation sources now opens up the way to provide more neutrons with improved cost and energy efficiency. A quantitative analysis of the figure-of-merit of the specialized instruments for pulsed source operation shows that up to 2 orders of magnitude intensity gains can be achieved in the next decade, with the advent of high power spallation sources. The first stations on this road, the MW class short pulse spallation sources SNS in the Usa (under commissioning), and J-PARC in Japan (under construction) will be followed by the 5 MW long pulse European Spallation Source (ESS). Further progress, that can be envisaged on the longer term, could amount to as much as another factor of 10 improvement. (author)

  2. Calibration of a D-T neutron generator

    International Nuclear Information System (INIS)

    Ito, Tadayuki

    1980-01-01

    The energy and production rate of neutrons from a thick target are discussed. The production rate of D-T neutrons is estimated by counting alpha particles with a silicon detector. In this case, it is necessary to evaluate a correction factor from the energy of deuteron, the reaction cross section, the stopping power of target materials and others. The factor was calculated and is shown in a figure. The energy spectrum of emitted neutrons is also estimated, where the atomic ratio of T and Ti is taken as a parameter. The shape of the spectrum is determined by the reaction cross section, and is not dependent on the ratio T/Ti. The errors due to competitive reactions, such as D(d, n) and D(d, p), are negligible. It is necessary for mutual comparison to take care of the target thickness, the acceleration voltage of D beam, the alpha-detector position, and the gain fluctuation of electronic circuits. (Kato, T.)

  3. Monte Carlo efficiency calibration of a neutron generator-based total-body irradiator

    International Nuclear Information System (INIS)

    Shypailo, R.J.; Ellis, K.J.

    2009-01-01

    Many body composition measurement systems are calibrated against a single-sized reference phantom. Prompt-gamma neutron activation (PGNA) provides the only direct measure of total body nitrogen (TBN), an index of the body's lean tissue mass. In PGNA systems, body size influences neutron flux attenuation, induced gamma signal distribution, and counting efficiency. Thus, calibration based on a single-sized phantom could result in inaccurate TBN values. We used Monte Carlo simulations (MCNP-5; Los Alamos National Laboratory) in order to map a system's response to the range of body weights (65-160 kg) and body fat distributions (25-60%) in obese humans. Calibration curves were constructed to derive body-size correction factors relative to a standard reference phantom, providing customized adjustments to account for differences in body habitus of obese adults. The use of MCNP-generated calibration curves should allow for a better estimate of the true changes in lean tissue mass that many occur during intervention programs focused only on weight loss. (author)

  4. Responsibility and Generativity in Online Learning Communities

    Science.gov (United States)

    Beth, Alicia D.; Jordan, Michelle E.; Schallert, Diane L.; Reed, JoyLynn H.; Kim, Minseong

    2015-01-01

    The purpose of this study was to investigate whether and how students enact "responsibility" and "generativity" through their comments in asynchronous online discussions. "Responsibility" referred to discourse markers indicating participants' sense that their contributions are required in order to uphold their…

  5. Results of the systematic study of neutron dosimeters and neutron radiometers responses from Bruyeres and Valduc's experiments

    International Nuclear Information System (INIS)

    Pras, Ph.; Ledoux, X.; Patin, Y.

    1999-01-01

    This document gives the results of the systematic study of neutron dosimeters (Bubbles detectors) and neutron radiometers (Cramal, Nausicaa, EGG Lb6411) with standard sources and with the Bruyeres Van de Graaff 4MV accelerator. The dose equivalent rate response as a function of the neutron energy is parameterized. Even for low dose equivalent rate, a good reproducibility of the measurements is found in the strict respect of a given method. The response of the different systems is independent of the dose equivalent rate. (author)

  6. Application of probability generating function to the essentials of nondestructive nuclear materials assay system using neutron correlation

    International Nuclear Information System (INIS)

    Hosoma, Takashi

    2017-01-01

    In the previous research (JAEA-Research 2015-009), essentials of neutron multiplicity counting mathematics were reconsidered where experiences obtained at the Plutonium Conversion Development Facility were taken into, and formulae of multiplicity distribution were algebraically derived up to septuplet using a probability generating function to make a strategic move in the future. Its principle was reported by K. Böhnel in 1985, but such a high-order expansion was the first case due to its increasing complexity. In this research, characteristics of the high-order correlation were investigated. It was found that higher-order correlation increases rapidly in response to the increase of leakage multiplication, crosses and leaves lower-order correlations behind, when leakage multiplication is > 1.3 that depends on detector efficiency and counter setting. In addition, fission rates and doubles count rates by fast neutron and by thermal neutron in their coexisting system were algebraically derived using a probability generating function again. Its principle was reported by I. Pázsit and L. Pál in 2012, but such a physical interpretation, i.e. associating their stochastic variables with fission rate, doubles count rate and leakage multiplication, is the first case. From Rossi-alpha combined distribution and measured ratio of each area obtained by Differential Die-Away Self-Interrogation (DDSI) and conventional assay data, it is possible to estimate: the number of induced fissions per unit time by fast neutron and by thermal neutron; the number of induced fissions (< 1) by one source neutron; and individual doubles count rates. During the research, a hypothesis introduced in their report was proved to be true. Provisional calculations were done for UO_2 of 1∼10 kgU containing ∼ 0.009 wt% "2"4"4Cm. (author)

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

    Science.gov (United States)

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

    2017-07-01

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

  8. Neutron generator for BNCT based on high current ECR ion source with gyrotron plasma heating.

    Science.gov (United States)

    Skalyga, V; Izotov, I; Golubev, S; Razin, S; Sidorov, A; Maslennikova, A; Volovecky, A; Kalvas, T; Koivisto, H; Tarvainen, O

    2015-12-01

    BNCT development nowadays is constrained by a progress in neutron sources design. Creation of a cheap and compact intense neutron source would significantly simplify trial treatments avoiding use of expensive and complicated nuclear reactors and accelerators. D-D or D-T neutron generator is one of alternative types of such sources for. A so-called high current quasi-gasdynamic ECR ion source with plasma heating by millimeter wave gyrotron radiation is suggested to be used in a scheme of D-D neutron generator in the present work. Ion source of that type was developed in the Institute of Applied Physics of Russian Academy of Sciences (Nizhny Novgorod, Russia). It can produce deuteron ion beams with current density up to 700-800 mA/cm(2). Generation of the neutron flux with density at the level of 7-8·10(10) s(-1) cm(-2) at the target surface could be obtained in case of TiD2 target bombardment with deuteron beam accelerated to 100 keV. Estimations show that it is enough for formation of epithermal neutron flux with density higher than 10(9) s(-1) cm(-2) suitable for BNCT. Important advantage of described approach is absence of Tritium in the scheme. First experiments performed in pulsed regime with 300 mA, 45 kV deuteron beam directed to D2O target demonstrated 10(9) s(-1) neutron flux. This value corresponds to theoretical estimations and proofs prospects of neutron generator development based on high current quasi-gasdynamic ECR ion source. Copyright © 2015 Elsevier Ltd. All rights reserved.

  9. Determination of the emission rate for the 14 MeV neutron generator with the use of radio-yttrium

    Directory of Open Access Journals (Sweden)

    Laszynska Ewa

    2015-06-01

    Full Text Available The neutron emission rate is a crucial parameter for most of the radiation sources that emit neutrons. In the case of large fusion devices the determination of this parameter is necessary for a proper assessment of the power release and the prediction for the neutron budget. The 14 MeV neutron generator will be used for calibration of neutron diagnostics at JET and ITER facilities. The stability of the neutron generator working parameters like emission and angular homogeneity affects the accuracy of calibration other neutron diagnostics. The aim of our experiment was to confirm the usefulness of yttrium activation method for monitoring of the neutron generator SODERN Model: GENIE 16. The reaction rate induced by neutrons inside the yttrium sample was indirectly measured by activation of the yttrium sample, and then by means of the γ-spectrometry method. The pre-calibrated HPGe detector was used to determine the yttrium radioactivity. The emissivity of neutron generator calculated on the basis of the measured radioactivity was compared with the value resulting from its electrical settings, and both of these values were found to be consistent. This allowed for a positive verification of the reaction cross section that was used to determine the reaction rate (6.45 × 10−21 reactions per second and the neutron emission rate (1.04 × 108 n·s−1. Our study confirms usefulness of the yttrium activation method for monitoring of the neutron generator.

  10. A measurement of the response to fast neutrons of several materials dosemeters

    International Nuclear Information System (INIS)

    Jones, L.T.; Kitching, S.J.; Lewis, T.A.; Playle, T.S.

    1986-07-01

    The response to fast neutrons was measured for three types of materials testing dosemeters: fast neutron dosimetry silicon diodes; beryllia, alumina and calcium fluoride TLDs; graphite walled ionisation chambers. The calibrations were made using a 3MW positive ion accelerator. The arrangement of the target, beam monitor and devices is described, and the measured fast neutron sensitivities are presented. (UK)

  11. 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)

  12. Next Generation Neutron Scintillators Based On Semiconductor Nanostructures

    International Nuclear Information System (INIS)

    Wang, Cai-Lin

    2008-01-01

    The results reported here successfully demonstrate the technical feasibility of ZnS QDs/ 6 LiF/polymer composites as thermal neutron scintillators. PartTec has obtained stable ZnS QDs with a quantum yield of 17% induced by UV light, and light pulse decay lifetimes of 10-30 ns induced by both UV and neutrons. These lifetime values are much shorter than those of commercial ZnS microparticle and 6 Li-glass scintillators. Clear pulse height peaks induced by neutron irradiation were seen for PartTec's ZnS nanocomposites. By adjusting the concentrations, particle size and degree of dispersion of ZnS QD/ 6 LiF in a PVA matrix, the light absorption and light yield of films at 420-440 nm can be optimized. PartTec's novel scintillators will replace traditional 6 Li-glass and ZnS/ 6 LiF:Ag scintillators if the PL quantum yield can be improved above 30%, and/or increase the transparency of present nanoscintillators. Time and resources inhibited PartTec's total success in Phase I. For example, bulk doping preparations of ZnS QDs with Ag + , Eu 3+ or Ce 3+ QDs was impractical given those constraints, nor did they permit PartTec to measure systematically the change of PL decay lifetimes in different samples. PartTec will pursue these studies in the current proposal, as well as develop a better capping and dopant along with developing brighter and faster ZnS QD scintillators.

  13. A compact neutron generator using a field ionization source.

    Science.gov (United States)

    Persaud, Arun; Waldmann, Ole; Kapadia, Rehan; Takei, Kuniharu; Javey, Ali; Schenkel, Thomas

    2012-02-01

    Field ionization as a means to create ions for compact and rugged neutron sources is pursued. Arrays of carbon nano-fibers promise the high field-enhancement factors required for efficient field ionization. We report on the fabrication of arrays of field emitters with a density up to 10(6) tips∕cm(2) and measure their performance characteristics using electron field emission. The critical issue of uniformity is discussed, as are efforts towards coating the nano-fibers to enhance their lifetime and surface properties.

  14. Neutron dosemeter responses in workplace fields and the implications of using realistic neutron calibration fields

    International Nuclear Information System (INIS)

    Thomas, D.J.; Horwood, N.; Taylor, G.C.

    1999-01-01

    The use of realistic neutron calibration fields to overcome some of the problems associated with the response functions of presently available dosemeters, both area survey instruments and personal dosemeters, has been investigated. Realistic calibration fields have spectra which, compared to conventional radionuclide source based calibration fields, more closely match those of the workplace fields in which dosemeters are used. Monte Carlo simulations were performed to identify laboratory systems which would produce appropriate workplace-like calibration fields. A detailed analysis was then undertaken of the predicted under- and over-responses of dosemeters in a wide selection of measured workplace field spectra assuming calibration in a selection of calibration fields. These included both conventional radionuclide source calibration fields, and also several proposed realistic calibration fields. The present state of the art for dosemeter performance, and the possibilities of improving accuracy by using realistic calibration fields are both presented. (author)

  15. Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

    2015-05-01

    Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. 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 fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. 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{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

  16. Desain Beam Shaping Assembly (BSA berbasis D-D Neutron Generator 2,45 MeV untuk Uji Fasilitas BNCT

    Directory of Open Access Journals (Sweden)

    Desman P. Gulo

    2015-12-01

    Full Text Available Boron Neutron Capture Therapy (BNCT is one of the cancer treatments that are being developed in nowadays. In order to support BNCT treatment for cancer that exists in underneath skin like breast cancer, the facility needs a generator that is able to produce epithermal neutron. One of the generator that is able to produce neutron is D-D neutron generator with 2.45 MeV energy. Based on the calculation of this paper, we found that the total production of neutron per second (neutron yield from Neutron Generator (NG by PSTA-BATAN Yogyakarta is 2.55×1011 n/s. The energy and flux that we found is in the range of quick neutron. Thus, it needs to be moderated to the level of epithermal neutron which is located in the interval energy of 1 eV to 10 KeV with 109 n/cm2s flux. This number is the recommendation standard from IAEA. Beam Shaping Assembly (BSA is needed in order to moderate the quick neutron to the level of epithermal neutron. One part of BSA that has the responsibility in moderating the quick neutron to epithermal neutron is the moderator. The substance of moderator used in this paper is MgF2 and A1F3. The thickness of moderator has been set in in such a way by using MCNPX software in order to fulfill the standard of IAEA. As the result of optimizing BSA moderator, the data obtain epithermal flux with the total number of 4.64×108 n/cm2/s for both of moderators with the thickness of moderator up to 15 cm. At the end of this research, the number of epithermal flux does not follow the standard of IAEA. This is because the flux neutron that is being produced by NG is relatively small. In conclusion, the NG from PSTA-BATAN Yogyakarta is not ready to be used for the BNCT treatment facility for the underneath skin cancer like breast cancer.

  17. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Tooth enamel dosimetric response to 2.8 MeV neutrons

    Science.gov (United States)

    Fattibene, P.; Angelone, M.; Pillon, M.; De Coste, V.

    2003-03-01

    Tooth enamel dosimetry, based on electron paramagnetic resonance (EPR) spectroscopy, is recognized as a powerful method for individual retrospective dose assessment. The method is mainly used for individual dose reconstruction in the epidemiological studies aimed at the radiation risk analysis. The study of the sensitivity of tooth enamel as a function of radiation quality is one of the main goals of the research in this field. In the present work, tooth enamel dose response in a monoenergetic neutron flux of 2.8 MeV, generated by the D-D reaction, was studied for in air and in phantom irradiations of enamel samples and of whole teeth. EPR measurements were complemented by Monte Carlo calculation and by gamma dose discrimination obtained with thermoluminescent and Geiger-Muller tube measurements. The 2.8 MeV neutrons to 60Co relative sensitivity was 0.33±0.08.

  19. Tooth enamel dosimetric response to 2.8 MeV neutrons

    International Nuclear Information System (INIS)

    Fattibene, P.; Angelone, M.; Pillon, M.; De Coste, V.

    2003-01-01

    Tooth enamel dosimetry, based on electron paramagnetic resonance (EPR) spectroscopy, is recognized as a powerful method for individual retrospective dose assessment. The method is mainly used for individual dose reconstruction in the epidemiological studies aimed at the radiation risk analysis. The study of the sensitivity of tooth enamel as a function of radiation quality is one of the main goals of the research in this field. In the present work, tooth enamel dose response in a monoenergetic neutron flux of 2.8 MeV, generated by the D-D reaction, was studied for in air and in phantom irradiations of enamel samples and of whole teeth. EPR measurements were complemented by Monte Carlo calculation and by gamma dose discrimination obtained with thermoluminescent and Geiger-Muller tube measurements. The 2.8 MeV neutrons to 60 Co relative sensitivity was 0.33±0.08

  20. Response of a carbon-walled proportional counter to 14 MeV neutrons

    International Nuclear Information System (INIS)

    Lewis, K.D.

    1982-01-01

    The response of a carbon-walled spherical proportional counter filled with a methane-based tissue-equivalent gas mixture at low pressure and irradiated with 14 MeV neutrons is first measured experimentally and is then calculated theoretically by using an analytical model. The model, called the CISS model, is derived from a consideration of four basic modes of interaction of charged particles generated in neutron-nucleus reactions with the spherical cavity of the detector. Since several quantities which have application in neutron dosimetry, radiation protection, and radiation biology make direct use of such spectra, it is desirable to have the ability to theoretically predict what is expected experimentally. Thus, a comparison between the two response curves is made. The discrepancy between them is investigated by considering several physical phenomena occurring within the detector wall which tend to distort the experimental response curve. In particular, the C(n,n',3α) reaction occurring in the detector wall gives rise to multiple events, originating from a single neutron interaction in the wall simultaneously strike the detector cavity, and are recorded as a single larger event in an experimental spectra. In the analytic model, the simultaneous entry of two charged particles into the cavity is scored as two separate smaller events, uncorrelated in their production. In this work, an effort is made to modify the analytic model prediction of the response curve by correcting for the multiple events which occur. Finally, the CISS model is used to compute mass stopping power corrections for this inhomogeneous detector

  1. Neutron Generation by Laser-Driven Spherically Convergent Plasma Fusion

    Science.gov (United States)

    Ren, G.; Yan, J.; Liu, J.; Lan, K.; Chen, Y. H.; Huo, W. Y.; Fan, Z.; Zhang, X.; Zheng, J.; Chen, Z.; Jiang, W.; Chen, L.; Tang, Q.; Yuan, Z.; Wang, F.; Jiang, S.; Ding, Y.; Zhang, W.; He, X. T.

    2017-04-01

    We investigate a new laser-driven spherically convergent plasma fusion scheme (SCPF) that can produce thermonuclear neutrons stably and efficiently. In the SCPF scheme, laser beams of nanosecond pulse duration and 1 014- 1 015 W /cm2 intensity uniformly irradiate the fuel layer lined inside a spherical hohlraum. The fuel layer is ablated and heated to expand inwards. Eventually, the hot fuel plasmas converge, collide, merge, and stagnate at the central region, converting most of their kinetic energy to internal energy, forming a thermonuclear fusion fireball. With the assumptions of steady ablation and adiabatic expansion, we theoretically predict the neutron yield Yn to be related to the laser energy EL, the hohlraum radius Rh, and the pulse duration τ through a scaling law of Yn∝(EL/Rh1.2τ0.2 )2.5. We have done experiments at the ShengGuangIII-prototype facility to demonstrate the principle of the SCPF scheme. Some important implications are discussed.

  2. Sealed operation of a rf driven ion source for a compact neutron generator to be used for associated particle imaging.

    Science.gov (United States)

    Wu, Y; Hurley, J P; Ji, Q; Kwan, J W; Leung, K N

    2010-02-01

    We present the recent development of a prototype compact neutron generator to be used in conjunction with the method of associated particle imaging for the purpose of active neutron interrogation. In this paper, the performance and device specifications of these compact generators that employ rf driven ion sources will be discussed. Initial measurements of the generator performance include a beam spot size of 1 mm in diameter and a neutron yield of 2x10(5) n/s with air cooling.

  3. Overview of the Division 2351 Neutron Generator Test Facility waveform digitizing system

    International Nuclear Information System (INIS)

    Bryant, T.C. Jr.

    1978-02-01

    All neutron generator waveforms from units tested at the SLA neutron generator test site are digitized and the digitized data stored in the CDC 6600 tape library for display and analysis using the CDC 6600 computer. The digitizing equipment consists mainly of seven Biomation Model 8100 transient recorders, Digital Equipment Corporation PDP 11/20 computer, RK05 disk, seven-track magnetic tape transport, and appropriate DEC and SLA controllers and interfaces. The PDP 11/20 computer is programmed in BASIC with assembly language drivers. In addition to digitizing waveforms, this equipment is used for other functions such as the automated testing of multiple-operation electronic neutron generators. Although other types of analysis have been done, the largest use of the digitized data has been for various types of graphical displays using the CDC 6600 and either the SD4020 or DX4460 plotters

  4. Perspectives of development of linac-driver for the ITEP neutron generator

    International Nuclear Information System (INIS)

    Kozodaev, A.M.; Vengrov, R.M.; Drozdovskij, A.A.; Kolomiets, A.A.; Orlov, Yu.G.; Raskopin, A.M.; Skachkov, V.S.; Shvedov, O.V.

    1999-01-01

    The perspectives of developing the experimental accelerator-driven neutron generator being made in ITEP are discussed. The ITEP ADS neutron generator consists of the target-blanket assembly and the linear proton accelerator Istra-36. It is projected to introduce superconducting sections in the composition of the neutron generator linac-driven. The application of superconducting resonators allows to increase the particle energy up to 53 MeV at the average beam current 500 μA. The variants of raising the average current up to 5 mA by increasing the HF-system power are considered. The application of magnetohard materials permits to decrease the cost of the bend magnet and its dimensions. To improve the radiation situation it is proposed to use the graphite absorbers of particles [ru

  5. New generation of cryogen free advanced superconducting magnets for neutron scattering experiments

    International Nuclear Information System (INIS)

    Kirichek, O; Adroja, D T; Manuel, P; Bewley, R I; Brown, J; Kouzmenko, G; Wotherspoon, R

    2012-01-01

    Recent advances in superconducting technology and cryocooler refrigeration have resulted in a new generation of advanced superconducting magnets for neutron beam applications. These magnets have outstanding parameters such as high homogeneity and stability at highest magnetic fields possible, a reasonably small stray field, low neutron scattering background and larger exposure to neutron detectors. At the same time the pulse tube refrigeration technology provides a complete re-condensing regime which allows to minimise the requirements for cryogens without introducing additional noise and mechanical vibrations. The magnets can be used with dilution refrigerator insert which expands the temperature range from 20mK to 300K. Here we are going to present design, test results and the operational data of the 14T magnet for neutron diffraction and the 9T wide angle chopper magnet for neutron spectroscopy developed by Oxford Instruments in collaboration with ISIS neutron source. First scientific results obtained from the neutron scattering experiments with these magnets are also going to be discussed.

  6. Massively parallel performance of neutron transport response matrix algorithms

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  7. Aerial Neutron Detection: Neutron Signatures for Nonproliferation and Emergency Response Applications

    Energy Technology Data Exchange (ETDEWEB)

    Maurer, Richard J.; Stampahar, Thomas G.; Smith, Ethan X.; Mukhopadhyay, Sanjoy; Wolff, Ronald S.; Rourke, Timothy J.; LeDonne, Jeffrey P.; Avaro, Emanuele; Butler, D. Andre; Borders, Kevin L.; Stampahar, Jezabel; Schuck, William H.; Selfridge, Thomas L.; McKissack, Thomas M.; Duncan, William W.; Hendricks, Thane J.

    2012-10-17

    From 2007 to the present, the Remote Sensing Laboratory has been conducting a series of studies designed to expand our fundamental understanding of aerial neutron detection with the goal of designing an enhanced sensitivity detection system for long range neutron detection. Over 35 hours of aerial measurements in a helicopter were conducted for a variety of neutron emitters such as neutron point sources, a commercial nuclear power reactor, nuclear reactor spent fuel in dry cask storage, depleted uranium hexafluoride and depleted uranium metal. The goals of the project were to increase the detection sensitivity of our instruments such that a 5.4 × 104 neutron/second source could be detected at 100 feet above ground level at a speed of 70 knots and to enhance the long-range detection sensitivity for larger neutron sources, i.e., detection ranges above 1000 feet. In order to increase the sensitivity of aerial neutron detection instruments, it is important to understand the dynamics of the neutron background as a function of altitude. For aerial neutron detection, studies have shown that the neutron background primarily originates from above the aircraft, being produced in the upper atmosphere by galactic cosmic-ray interactions with air molecules. These interactions produce energetic neutrons and charged particles that cascade to the earth’s surface, producing additional neutrons in secondary collisions. Hence, the neutron background increases as a function of altitude which is an impediment to long-range neutron detection. In order to increase the sensitivity for long range detection, it is necessary to maintain a low neutron background as a function of altitude. Initial investigations show the variation in the neutron background can be decreased with the application of a cosmic-ray shield. The results of the studies along with a representative data set are presented.

  8. Laser driven compression and neutron generation with spherical shell targets

    International Nuclear Information System (INIS)

    Campbell, P.M.; Hammerling, P.; Johnson, R.R.; Kubis, J.J.; Mayer, F.J.

    1977-01-01

    Laser-driven implosion experiments using DT-gas-filled spherical glass-shell targets are described. Neutron yields to 5 x 10 7 are produced from implosions of small ( -- 55 μm-diameter) targets spherically illuminated with an on-target laser power of 0.4 terawatt. Nuclear reaction product diagnostics, X-ray pinhole photographs, fast-ion spectra and X-ray measurements are used in conjunction with hydrodynamic computer code simulations to investigate the implosion phenomenology as well as the target corona evolution. Simulations using completely classical effects are not able to describe the full range of experimental data. Electron or radiation preheating may be required to explain some implosion measurements. (auth.)

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

    CERN Document Server

    Rehm, C

    2002-01-01

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

  10. The intense neutron generator INGE-1 at the Technical University of Dresden

    International Nuclear Information System (INIS)

    Bittner, M.; Meisner, A.; Paffrath, E.; Schwiers, H.; Seeliger, D.

    1989-01-01

    The INGE-1 neutron generator developed for intergal 14 MeV neutron experiments is described. The accelerator produces steady d + ion beam in the current range of 1-10 mA with 120-240 keV energies at the target position. The beam is produced with a combined duoplasmatron focalization system on high voltage. A 30 keV beam is accelerated on final energy by a two-gap acceleration tube. The estimations conducted show that the generator maximum strength can reach up to 2x10 12 s -1 at 10 mA beam current and 220 keV energy. 4 refs

  11. High-Flux Neutron Generator Facility for Geochronology and Nuclear Physics Research

    Science.gov (United States)

    Waltz, Cory; HFNG Collaboration

    2015-04-01

    A facility based on a next-generation, high-flux D-D neutron generator (HFNG) is being commissioned at UC Berkeley. The generator is designed to produce monoenergetic 2.45 MeV neutrons at outputs exceeding 1011 n/s. The HFNG is designed around two RF-driven multi-cusp ion sources that straddle a titanium-coated copper target. D + ions, accelerated up to 150 keV from the ion sources, self-load the target and drive neutron generation through the d(d,n)3 He fusion reaction. A well-integrated cooling system is capable of handling beam power reaching 120 kW impinging on the target. The unique design of the HFNG target permits experimental samples to be placed inside the target volume, allowing the samples to receive the highest neutron flux (1011 cm-2 s-1) possible from the generator. In addition, external beams of neutrons will be available simultaneously, ranging from thermal to 2.45 MeV. Achieving the highest neutron yields required carefully designed schemes to mitigate back-streaming of high energy electrons liberated from the cathode target by deuteron bombardment. The proposed science program is focused on pioneering advances in the 40 Ar/39 Ar dating technique for geochronology, new nuclear data measurements, basic nuclear science, and education. An end goal is to become a user facility for researchers. This work is supported by NSF Grant No. EAR-0960138, U.S. DOE LBNL Contract No. DE-AC02-05CH11231, U.S. DOE LLNL Contract No. DE-AC52-07NA27344, and UC Office of the President Award 12-LR-238745.

  12. Utilization of a sealed-tube neutron generator for training and research

    International Nuclear Information System (INIS)

    Jonah, S.A.

    2000-01-01

    The development of a program in nuclear science and technology in Nigeria began in 1976 with the establishment of two research centers, namely, the Centre for Energy Research and Training, (CERT), Zaria and the Centre for Energy Research and Development (CERD), Ile-Ife. The choice of Neutron Activation Analysis (NAA) technique as a very effective method of training scientists in basic and applied nuclear research led to the purchase of two KAMAN A-711 Neutron Generators for the two research centers. At CERT, the neutron generator (code named ZARABUNG-1) was successfully installed and the first 14 MeV neutrons were produced through the technical assistance of the International Atomic Energy Agency (IAEA) in 1988. In 1991, a new tube-head was purchased and installed due to the expiration of the old tube. Following the completion of its permanent site, the neutron generator was re-located from the old site and re-installed at the permanent site of CERT in 1995. (author)

  13. Detector Response to Neutrons Slowed Down in Media Containing Cadmium

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1943-07-01

    This report was written by E. Broda, H. Hereward and L. Kowarski at the Cavendish Laboratory (Cambridge) in September 1943 and is about the detector response to neutrons slowed down in media containing cadmium. The following measurement description and the corresponding results can be found in this report: B, Mn, In, I, Dy and Ir detectors were activated, with and without a Cd shield, near the source in a vessel containing 7 litres of water or solutions of CdSO{sub 4} ranging between 0.1 and 2.8 mols per litre. Numerical data on observed activities are discussed in two different ways and the following conclusions can be drawn: The capture cross-section of dysprosium decreases quicker than 1/v and this discrepancy becomes noticeable well within the limits of the C-group. This imposes obvious limitations on the use of Dy as a detector of thermal neutrons. Cadmium differences of manganese seem to be a reliable 1/v detector for the whole C-group. Indium and iridium show definite signs of an increase of vσ in the upper regions of the C-group. Deviations shown by iodine are due to the imperfections of the technique rather than to a definite departure from the 1/v law. (nowak)

  14. Response of combined albedo-track neutron personnel dosimeters behind IHEP proton synchrotron shielding

    International Nuclear Information System (INIS)

    Sannikov, A.V.; Korshunova, E.P.

    1989-01-01

    The method of readings interpretation of combined albedo-track neutron personnel dosemeters based on calculationsl analysis of the detector responses in various neutron spectra is described. The measurements of dose equivalent responses have been performed in various points behind IHEP proton synchrotron shielding. It is shown that CDs with fission track detectors have a small dose equivalent response dispersion behind IHEP proton synchrotron shielding, that shows the promise of their using for neutron personnel monitoring, that shows the promise of their using for neutron personnel monitoring at high energy accelerators. 16 refs.; 7 figs.; 3 tabs

  15. The side-on response of a standard long counter to fast neutrons

    International Nuclear Information System (INIS)

    Johnson, F.A.

    1979-01-01

    The response of a standard long counter to neutrons incident into its front face relative to its response to those incident into its side was measured for a range of neutron energies, and an increasing sensitivity to high-energy neutrons incident into the side was evident. The effect of a shadow bar in contributing to an initial degradation in energy of neutrons which then scatter from the surroundings into the counter was suggested by the response of the counter in the side-on orientation as a function of the separation distance of the bar from a source. (Auth.)

  16. Nitrogen Detection in Bulk Samples Using a D-D Reaction-Based Portable Neutron Generator

    Directory of Open Access Journals (Sweden)

    A. A. Naqvi

    2013-01-01

    Full Text Available Nitrogen concentration was measured via 2.52 MeV nitrogen gamma ray from melamine, caffeine, urea, and disperse orange bulk samples using a newly designed D-D portable neutron generator-based prompt gamma ray setup. Inspite of low flux of thermal neutrons produced by D-D reaction-based portable neutron generator and interference of 2.52 MeV gamma rays from nitrogen in bulk samples with 2.50 MeV gamma ray from bismuth in BGO detector material, an excellent agreement between the experimental and calculated yields of nitrogen gamma rays indicates satisfactory performance of the setup for detection of nitrogen in bulk samples.

  17. Study of the response of a lithium yttrium borate scintillator based neutron rem counter by Monte Carlo radiation transport simulations

    Science.gov (United States)

    Sunil, C.; Tyagi, Mohit; Biju, K.; Shanbhag, A. A.; Bandyopadhyay, T.

    2015-12-01

    The scarcity and the high cost of 3He has spurred the use of various detectors for neutron monitoring. A new lithium yttrium borate scintillator developed in BARC has been studied for its use in a neutron rem counter. The scintillator is made of natural lithium and boron, and the yield of reaction products that will generate a signal in a real time detector has been studied by FLUKA Monte Carlo radiation transport code. A 2 cm lead introduced to enhance the gamma rejection shows no appreciable change in the shape of the fluence response or in the yield of reaction products. The fluence response when normalized at the average energy of an Am-Be neutron source shows promise of being used as rem counter.

  18. Study of the response of a lithium yttrium borate scintillator based neutron rem counter by Monte Carlo radiation transport simulations

    Energy Technology Data Exchange (ETDEWEB)

    Sunil, C., E-mail: csunil11@gmail.com [Accelerator Radiation Safety Section, Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tyagi, Mohit [Technical Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Biju, K.; Shanbhag, A.A.; Bandyopadhyay, T. [Accelerator Radiation Safety Section, Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2015-12-11

    The scarcity and the high cost of {sup 3}He has spurred the use of various detectors for neutron monitoring. A new lithium yttrium borate scintillator developed in BARC has been studied for its use in a neutron rem counter. The scintillator is made of natural lithium and boron, and the yield of reaction products that will generate a signal in a real time detector has been studied by FLUKA Monte Carlo radiation transport code. A 2 cm lead introduced to enhance the gamma rejection shows no appreciable change in the shape of the fluence response or in the yield of reaction products. The fluence response when normalized at the average energy of an Am–Be neutron source shows promise of being used as rem counter.

  19. SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response

    International Nuclear Information System (INIS)

    Furuta, Lazuo; Kondo, Shunsuke; Oka, Yoshika

    1991-01-01

    1 - Description of program or function: SUSD calculates sensitivity coefficients for one and two-dimensional transport problems. Variance and standard deviation of detector responses or design parameters can be obtained using cross-section covariance matrices. In neutron transport problems, this code is able to perform sensitivity-uncertainty analysis for secondary angular distribution (SAD) or secondary energy distribution (SED). 2 - Method of solution: The first-order perturbation theory is used to obtain sensitivity coefficients. The method described in the distributed report is employed to consider SAD/SED effect. 3 - Restrictions on the complexity of the problem: Variable dimension is used so that there is no limitation in each array size but the total core size

  20. Response of CsI:Pb Scintillator Crystal to Neutron Radiation

    Science.gov (United States)

    Costa Pereira, Maria da Conceição; Filho, Tufic Madi; Berretta, José Roberto; Náhuel Cárdenas, José Patrício; Iglesias Rodrigues, Antonio Carlos

    2018-01-01

    The helium-3 world crisis requires a development of new methods of neutron detection to replace commonly used 3He proportional counters. In the past decades, great effort was made to developed efficient and fast scintillators to detect radiation. The inorganic scintillator may be an alternative. Inorganic scintillators with much higher density should be selected for optimal neutron detection efficiency taking into consideration the relevant reactions leading to light emission. These detectors should, then, be carefully characterized both experimentally and by means of advanced simulation code. Ideally, the detector should have the capability to separate neutron and gamma induced events either by amplitude or through pulse shape differences. As neutron sources also generate gamma radiation, which can interfere with the measurement, it is necessary that the detector be able to discriminate the presence of such radiation. Considerable progress has been achieved to develop new inorganic scintillators, in particular increasing the light output and decreasing the decay time by optimized doping. Crystals may be found to suit neutron detection. In this report, we will present the results of the study of lead doped cesium iodide crystals (CsI:Pb) grown in our laboratory, using the vertical Bridgman technique. The concentration of the lead doping element (Pb) was studied in the range 5x10-4 M to 10-2 M . The crystals grown were subjected to annealing (heat treatment). In this procedure, vacuum of 10-6 mbar and continuous temperature of 350°C, for 24 hours, were employed. In response to neutron radiation, an AmBe source with energy range of 1 MeV to 12 MeV was used. The activity of the AmBe source was 1Ci Am. The fluency was 2.6 x 106 neutrons/second. The operating voltage of the photomultiplier tube was 1700 V; the accumulation time in the counting process was 600 s and 1800 s. The scintillator crystals used were cut with dimensions of 20 mm diameter and 10 mm height.

  1. A novel design of beam shaping assembly to use D-T neutron generator for BNCT.

    Science.gov (United States)

    Kasesaz, Yaser; Karimi, Marjan

    2016-12-01

    In order to use 14.1MeV neutrons produced by d-T neutron generators, two special and novel Beam Shaping Assemblies (BSA), including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. The results show that the proposed BSA can provide the qualified epithermal neutron beam for BNCT. The final epithermal neutron flux is about 6e9 n/cm2.s. The final proposed BSA has some different advantages: 1) it consists of usual and well-known materials (Pb, Al, Fluental and Cd); 2) it has a simple geometry; 3) it does not need any additional gamma filter; 4) it can provide high flux of epithermal neutrons. As this type of neutron source is under development in the world, it seems that they can be used clinically in a hospital considering the proposed BSA. Copyright © 2016 Elsevier Ltd. All rights reserved.

  2. Staged Z-pinch Experiments at the 1MA Zebra pulsed-power generator: Neutron measurements

    Science.gov (United States)

    Ruskov, Emil; Darling, T.; Glebov, V.; Wessel, F. J.; Anderson, A.; Beg, F.; Conti, F.; Covington, A.; Dutra, E.; Narkis, J.; Rahman, H.; Ross, M.; Valenzuela, J.

    2017-10-01

    We report on neutron measurements from the latest Staged Z-pinch experiments at the 1MA Zebra pulsed-power generator. In these experiments a hollow shell of argon or krypton gas liner, injected between the 1 cm anode-cathode gap, compresses a deuterium plasma target of varying density. Axial magnetic field Bz neutron Time of Flight (nTOF) detectors are augmented with a large area ( 1400 cm2) liquid scintillator detector to which fast gatedPhotek photomultipliers are attached. Sample data from these neutron diagnostics systems is presented. Consistently high neutron yields YDD >109 are measured, with highest yield of 2.6 ×109 . A pair of horizontally and vertically placed plastic scintillator nTOFs suggest isotropic i.e. thermonuclear origin of the neutrons produced. nTOF data from the liquid scintillator detector was cross-calibrated with the silver activation detector, and can be used for accurate calculation of the neutron yield. Funded by the Advanced Research Projects Agency - Energy, under Grant Number DE-AR0000569.

  3. Sustaining knowledge in the neutron generator community and benchmarking study. Phase II.

    Energy Technology Data Exchange (ETDEWEB)

    Huff, Tameka B.; Stubblefield, William Anthony; Cole, Benjamin Holland, II; Baldonado, Esther

    2010-08-01

    This report documents the second phase of work under the Sustainable Knowledge Management (SKM) project for the Neutron Generator organization at Sandia National Laboratories. Previous work under this project is documented in SAND2008-1777, Sustaining Knowledge in the Neutron Generator Community and Benchmarking Study. Knowledge management (KM) systems are necessary to preserve critical knowledge within organizations. A successful KM program should focus on people and the process for sharing, capturing, and applying knowledge. The Neutron Generator organization is developing KM systems to ensure knowledge is not lost. A benchmarking study involving site visits to outside industry plus additional resource research was conducted during this phase of the SKM project. The findings presented in this report are recommendations for making an SKM program successful. The recommendations are activities that promote sharing, capturing, and applying knowledge. The benchmarking effort, including the site visits to Toyota and Halliburton, provided valuable information on how the SEA KM team could incorporate a KM solution for not just the neutron generators (NG) community but the entire laboratory. The laboratory needs a KM program that allows members of the workforce to access, share, analyze, manage, and apply knowledge. KM activities, such as communities of practice (COP) and sharing best practices, provide a solution towards creating an enabling environment for KM. As more and more people leave organizations through retirement and job transfer, the need to preserve knowledge is essential. Creating an environment for the effective use of knowledge is vital to achieving the laboratory's mission.

  4. Sustaining knowledge in the neutron generator community and benchmarking study. Phase II

    International Nuclear Information System (INIS)

    Huff, Tameka B.; Stubblefield, William Anthony; Cole, Benjamin Holland II; Baldonado, Esther

    2010-01-01

    This report documents the second phase of work under the Sustainable Knowledge Management (SKM) project for the Neutron Generator organization at Sandia National Laboratories. Previous work under this project is documented in SAND2008-1777, Sustaining Knowledge in the Neutron Generator Community and Benchmarking Study. Knowledge management (KM) systems are necessary to preserve critical knowledge within organizations. A successful KM program should focus on people and the process for sharing, capturing, and applying knowledge. The Neutron Generator organization is developing KM systems to ensure knowledge is not lost. A benchmarking study involving site visits to outside industry plus additional resource research was conducted during this phase of the SKM project. The findings presented in this report are recommendations for making an SKM program successful. The recommendations are activities that promote sharing, capturing, and applying knowledge. The benchmarking effort, including the site visits to Toyota and Halliburton, provided valuable information on how the SEA KM team could incorporate a KM solution for not just the neutron generators (NG) community but the entire laboratory. The laboratory needs a KM program that allows members of the workforce to access, share, analyze, manage, and apply knowledge. KM activities, such as communities of practice (COP) and sharing best practices, provide a solution towards creating an enabling environment for KM. As more and more people leave organizations through retirement and job transfer, the need to preserve knowledge is essential. Creating an environment for the effective use of knowledge is vital to achieving the laboratory's mission.

  5. Use of a pulsed neutron generator for in vivo measurement of body carbon

    International Nuclear Information System (INIS)

    Kehayias, J.J.; Ellis, K.J.; Cohn, S.H.; Yasumura, S.

    1986-01-01

    The measurement of total body fat is of importance in studies of nutritional assessment, dietary regimens, and for the management of obesity. In the past, fat has been determined either by anthropometric methods, which introduce high uncertainties, or by model-dependent estimation of fat-free tissue. The validity, however, of the different models in disease is questionable. Total body carbon measurements provide a more direct evaluation of body fat both in normal subjects and in patients. The authors present here a facility for carbon measurements without the use of a major accelerator. The same facility can be used for the measurement of other major body elements and for the evaluation of the body's compartments. Carbon is measured in vivo through neutron inelastic scattering, by detecting the 4.44 MeV gamma rays. A miniature (10 cm long) 14 MeV D-T neutron generator is used. The short half-life of the 4.44 MeV state of carbon requires detection of the gamma rays simultaneously with the 10 μs neutron pulse. Generators with low pulsing rate were found inappropriate for carbon measurement because of their low duty-cycle (high neutron output during pulse). The detection system consists of NaI(T1) detectors and fast electronics for handling the high even rate during the neutron pulse. A description of the facility and an evaluation of the technique will be presented

  6. Production of neutron-rich nuclei in fission induced by neutrons generated by the p+ sup 1 sup 3 C reaction at 55 MeV

    CERN Document Server

    Stroe, L; Andrighetto, A; Tecchio, L B; Dendooven, P; Huikari, J; Pentillä, H; Peraejaervi, K; Wang, Y

    2003-01-01

    Cross-sections for the production of neutron-rich nuclei obtained by neutron-induced fission of natural uranium have been measured. The neutrons were generated by bombarding a sup 1 sup 3 C target with 55 MeV protons. The results, position of the maximum in the (Z, A)-plane, width and magnitude, are very comparable with those where the neutrons are generated by bombardment of natural sup 1 sup 2 C graphite with 50 MeV deuterons. Depending on the geometry of the converter/target assembly the isotope yields, however, are a factor of 2-3 lower due to less efficient production of neutrons per primary projectile, especially at small forward angles. (orig.)

  7. Low-dose neutron dose response of zebrafish embryos obtained from the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility

    International Nuclear Information System (INIS)

    Ng, C.Y.P.; Kong, E.Y.; Konishi, T.; Kobayashi, A.; Suya, N.; Cheng, S.H.; Yu, K.N.

    2015-01-01

    The dose response of embryos of the zebrafish, Danio rerio, irradiated at 5 h post fertilization (hpf) by 2-MeV neutrons with ≤100 mGy was determined. The neutron irradiations were made at the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility in the National Institute of Radiological Sciences (NIRS), Chiba, Japan. A total of 10 neutron doses ranging from 0.6 to 100 mGy were employed (with a gamma-ray contribution of 14% to the total dose), and the biological effects were studied through quantification of apoptosis at 25 hpf. The responses for neutron doses of 10, 20, 25, and 50 mGy approximately fitted on a straight line, while those for neutron doses of 0.6, 1 and 2.5 mGy exhibited neutron hormetic effects. As such, hormetic responses were generically developed by different kinds of ionizing radiations with different linear energy transfer (LET) values. The responses for neutron doses of 70 and 100 mGy were significantly below the lower 95% confidence band of the best-fit line, which strongly suggested the presence of gamma-ray hormesis. - Highlights: • Neutron dose response was determined for embryos of the zebrafish, Danio rerio. • Neutron doses of 0.6, 1 and 2.5 mGy led to neutron hormetic effects. • Neutron doses of 70 and 100 mGy accompanied by gamma rays led to gamma-ray hormesis

  8. Development of response transforms from comparative study of commercial pulsed neutron capture logging systems

    International Nuclear Information System (INIS)

    Salaita, G.N.; Youngblood, W.E.

    1991-01-01

    This paper reports that the absence of a common calibration facility to ascertain the accuracy of commercial pulsed neutron capture logging systems, coupled with the desire for more accurate saturation determination from time-lapse logs, prompted Saudi Aramco to carry out this comparative study. Three generations of Schlumberger's Thermal Decay Time (TDT) logging devices, viz., TDT-K, TDT-M, and TDT-P along with Atlas Wireline PDK-100 system were run in an Aramco well. The wellbore 8-1/2 inch with 7-inch casing-penetrated clean sand, shaly sand, and shale streaks sequence as exhibited by the open hole natural gamma ray log. initially, the wellbore fluid was diesel. The fluid was then changed to brines of 42-kppm and 176-kppm NACl, respectively. Three repeat passes at a logging speed of 900 ft/hr were obtained by each device for each of the three borehole liquids. In the case of PDK-100, a second set of log runs was obtained at 1800 ft/hr. The results of this extensive comparative study have increased the author's understanding of the borehole liquid and the diffusion effects on the response of pulsed neutron capture logging systems and also on the relative accuracy and precision of measured formation thermal neutron capture cross section by each system

  9. Salient features, response and operation of Lead-Free Gulmarg Neutron Monitor

    International Nuclear Information System (INIS)

    Mufti, S.; Chatterjee, S.; Ishtiaq, P.M.; Darzi, M.A.; Mir, T.A.; Shah, G.N.

    2016-01-01

    Lead-Free Gulmarg Neutron Monitor (LFGNM) provides continuous ground level intensity measurements of atmospheric secondary neutrons produced in interactions of primary cosmic rays with the Earth's constituent atmosphere. We report the LFGNM detector salient features and simulation of its energy response for 10"−"1"1 MeV to 10"4 MeV energy incident neutrons using the FLUKA Monte Carlo package. An empirical calibration of the LFGNM detector carried out with a Pu–Be neutron source for maximising its few MeV neutron counting sensitivity is also presented. As an illustration of its functionality a single representative transient solar modulation event recorded by LFGNM depicting Forbush decrease in integrated neutron data for which the geospace consequences are well known is also presented. Performance of LFGNM under actual observation conditions for effectively responding to transient solar modulation is seen to compare well with other world-wide conventional neutron monitors.

  10. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    Science.gov (United States)

    Bowman, Charles D.

    1992-01-01

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  11. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    Science.gov (United States)

    Bowman, C.D.

    1992-11-03

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  12. Measurement and simulation of neutron response function of organic liquid scintillator detector

    International Nuclear Information System (INIS)

    Gohil, M.; Banerjee, K.; Bhattacharya, S.; Bhattacharya, C.; Kundu, S.; Rana, T.K.; Mukherjee, G.; Meena, J.K.; Pandey, R.; Pai, H.; Ghosh, T.K.; Dey, A.; Mukhopadhyay, S.; Pandit, D.; Pal, S.; Banerjee, S.R.; Bandhopadhyay, T.

    2012-01-01

    Response functions of monoenergetic neutrons at various energies, corresponding to a measured neutron energy spectrum have been extracted. The experimental response functions for neutron energies in the range of ∼2-20 MeV have been compared with the respective GEANT4 predictions. It has been found that, there is some discrepancy between the experimental and the GEANT4 simulated neutron response functions at lower pulse height regions, which increases with the increase of neutron energy. This might be due to the incompleteness of the physics processes used in the present GEANT4 simulations. In particular, higher order reaction processes which become more significant at higher energies should be properly taken into account in the calculation of response function.

  13. Copper benchmark experiment at the Frascati Neutron Generator for nuclear data validation

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M., E-mail: maurizio.angelone@enea.it; Flammini, D.; Loreti, S.; Moro, F.; Pillon, M.; Villari, R.

    2016-11-01

    Highlights: • A benchmark experiment was performed using pure copper with 14 MeV neutrons. • The experiment was performed at the Frascati Neutron Generator (FNG). • Activation foils, thermoluminescent dosimeters and scintillators were used to measure reactions rates (RR), nuclear heating and neutron spectra. • The paper presents the RR measurements and the post analysis using MCNP5 and JEFF-3.1.1, JEFF-3.2 and FENDL-3.1 libraries. • C/Es are presented showing the need for deep revision of Cu cross sections. - Abstract: A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 60 cm{sup 3}), aimed at testing and validating the recent nuclear data libraries for fusion applications, was performed at the 14-MeV Frascati Neutron Generator (FNG) as part of a F4E specific grant (F4E-FPA-395-01) assigned to the European Consortium on Nuclear Data and Experimental Techniques. The relevant neutronics quantities (e.g., reaction rates, neutron flux spectra, doses, etc.) were measured using different experimental techniques and the results were compared to the calculated quantities using fusion relevant nuclear data libraries. This paper focuses on the analyses carried-out by ENEA through the activation foils techniques. {sup 197}Au(n,γ){sup 198}Au, {sup 186}W(n,γ){sup 187}W, {sup 115}In(n,n′){sup 115}In, {sup 58}Ni(n,p){sup 58}Co, {sup 27}Al(n,α){sup 24}Na, {sup 93}Nb(n,2n){sup 92}Nb{sup m} activation reactions were used. The foils were placed at eight different positions along the Cu block and irradiated with 14 MeV neutrons. Activation measurements were performed by means of High Purity Germanium (HPGe) detector. Detailed simulation of the experiment was carried-out using MCNP5 Monte Carlo code and the European JEFF-3.1.1 and 3.2 nuclear cross-sections data files for neutron transport and IRDFF-v1.05 library for the reaction rates in activation foils. The calculated reaction rates (C) were compared to the experimental quantities (E) and

  14. Pulsed neutron generator system for astrobiological and geochemical exploration of planetary bodies

    International Nuclear Information System (INIS)

    Akkurt, Hatice; Groves, Joel L.; Trombka, Jacob; Starr, Richard; Evans, Larry; Floyd, Samuel; Hoover, Richard; Lim, Lucy; McClanahan, Timothy; James, Ralph; McCoy, Timothy; Schweitzer, Jeffrey

    2005-01-01

    A pulsed neutron/gamma-ray detection system for use on rovers to survey the elemental concentrations of Martian and Lunar surface and subsurface materials is evaluated. A robotic survey system combining a pulsed neutron generator (PNG) and detectors (gamma ray and neutron) can measure the major constituents to a depth of about 30 cm. Scanning mode measurements can give the major elemental concentrations while the rover is moving; analyzing mode measurements can give a detailed elemental analysis of the adjacent material when the rover is stationary. A detailed map of the subsurface elemental concentrations will provide invaluable information relevant to some of the most fundamental astrobiological questions including the presence of water, biogenic activity, life habitability and deposition processes

  15. Dynamical response of the nuclear 'pasta' in neutron star crusts

    International Nuclear Information System (INIS)

    Horowitz, C.J.; Perez-Garcia, M.A.; Berry, D.K.; Piekarewicz, J.

    2005-01-01

    The nuclear pasta - a novel state of matter having nucleons arranged in a variety of complex shapes - is expected to be found in the crust of neutron stars and in core-collapse supernovae at subnuclear densities of about 10 14 g/cm 3 . Owing to frustration, a phenomenon that emerges from the competition between short-range nuclear attraction and long-range Coulomb repulsion, the nuclear pasta displays a preponderance of unique low-energy excitations. These excitations could have a strong impact on many transport properties, such as neutrino propagation through stellar environments. The excitation spectrum of the nuclear pasta is computed via a molecular-dynamics simulation involving up to 100,000 nucleons. The dynamic response of the pasta displays a classical plasma oscillation in the 1- to 2-MeV region. In addition, substantial strength is found at low energies. Yet this low-energy strength is missing from a simple ion model containing a single-representative heavy nucleus. The low-energy strength observed in the dynamic response of the pasta is likely to be a density wave involving the internal degrees of freedom of the clusters

  16. Evaluation of energy responses for neutron dose-equivalent meters made in Japan

    International Nuclear Information System (INIS)

    Saegusa, J.; Yoshizawa, M.; Tanimura, Y.; Yoshida, M.; Yamano, T.; Nakaoka, H.

    2004-01-01

    Energy responses of three types of Japanese neutron dose-equivalent (DE) meters were evaluated by Monte Carlo simulations and measurements. The energy responses were evaluated for thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and also for neutrons from such radionuclide sources as 252 Cf and 241 Am-Be. The calculated results were corroborated with the measured ones. The angular dependence of the response and the DE response were also evaluated. As a result, reliable energy responses were obtained by careful simulations of the proportional counter, moderator and absorber of the DE meters. Furthermore, the relationship between pressure of counting gas and response of the DE meter was discussed. By using the obtained responses, relations between predicted readings of the DE meters and true DE values were studied for various workplace spectra

  17. Neutron response matrix for unfolding NE-213 measurements to 21 MeV

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Wehring, B.W.; Johnson, R.H.

    1976-01-01

    A neutron response matrix from measured neutron responses of NE-213 in the energy range of 0.2 to 22 MeV is presented. An interpolation scheme was used to construct an 81-column matrix from the data of Verbinski, Burrus, Love, Zobel, and Hill. As a test of the new response matrix, the Cf-252 neutron spectrum was measured and unfolded using the new response matrix and the FORIST unfolding code. The spectrum agrees well with previous measurements at lower energies, while providing new information above 8 MeV

  18. Large area imaging of hydrogenous materials using fast neutrons from a DD fusion generator

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, J.T., E-mail: ted@adelphitech.com [Adelphi Technology Inc., 2003 East Bayshore Road, Redwood City, California 94063 (United States); Williams, D.L.; Gary, C.K.; Piestrup, M.A.; Faber, D.R.; Fuller, M.J.; Vainionpaa, J.H.; Apodaca, M. [Adelphi Technology Inc., 2003 East Bayshore Road, Redwood City, California 94063 (United States); Pantell, R.H.; Feinstein, J. [Department of Electrical Engineering, Stanford University, Stanford, California 94305 (United States)

    2012-05-21

    A small-laboratory fast-neutron generator and a large area detector were used to image hydrogen-bearing materials. The overall image resolution of 2.5 mm was determined by a knife-edge measurement. Contact images of objects were obtained in 5-50 min exposures by placing them close to a plastic scintillator at distances of 1.5 to 3.2 m from the neutron source. The generator produces 10{sup 9} n/s from the DD fusion reaction at a small target. The combination of the DD-fusion generator and electronic camera permits both small laboratory and field-portable imaging of hydrogen-rich materials embedded in high density materials.

  19. High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

    International Nuclear Information System (INIS)

    Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.

    2011-01-01

    Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240 Pu . On the other hand, identification of shielded uranium requires active methods using neutron or photon sources . Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials . In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers . Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1x10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2x10 4 n/cm 2 s.

  20. Design of the Next Generation Target at the Lujan Neutron Scattering Center, LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Ferres, Laurent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); National Graduate School of Engineering and Research Center (ENSICAEN), Caen (France)

    2016-08-03

    Los Alamos National Laboratory (LANL) supports scientific research in many diverse fields such as biology, chemistry, and nuclear science. The Laboratory was established in 1943 during the Second World War to develop nuclear weapons. Today, LANL is one of the largest laboratories dedicated to nuclear defense and operates an 800 MeV proton linear accelerator for basic and applied research including: production of high- and low-energy neutrons beams, isotope production for medical applications and proton radiography. This accelerator is located at the Los Alamos Neutron Science Center (LANSCE). The work performed involved the redesign of the target for the low-energy neutron source at the Lujan Neutron Scattering Center, which is one of the facilities built around the accelerator. The redesign of the target involves modeling various arrangements of the moderator-reflector-shield for the next generation neutron production target. This is done using Monte Carlo N-Particle eXtended (MCNPX), and ROOT analysis framework, a C++ based-software, to analyze the results.

  1. Experimental characterization of the neutron spectra generated by a high-energy clinical LINAC

    Energy Technology Data Exchange (ETDEWEB)

    Amgarou, K., E-mail: khalil.amgarou@uab.e [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France); Lacoste, V.; Martin, A. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France)

    2011-02-11

    The production of unwanted neutrons by electron linear accelerators (LINACs) has attracted a special attention since the early 50s. The renewed interest in this topic during the last years is due mainly to the increased use of such machines in radiotherapy. Specially, in most of developing countries where many old teletherapy irradiators, based on {sup 60}Co and {sup 137}Cs radioactive sources, are being replaced with new LINAC units. The main objective of this work is to report the results of an experimental characterization of the neutron spectra generated by a high-energy clinical LINAC. Measurements were carried out, considering four irradiation configurations, by means of our recently developed passive Bonner sphere spectrometer (BSS) using pure gold activation foils as central detectors. This system offers the possibility to measure neutrons over a wide energy range (from thermal up to a few MeV) at pulsed, intense and complex mixed n-{gamma} fields. A two-step unfolding method that combines the NUBAY and MAXED codes was applied to derive the final neutron spectra as well as their associated integral quantities (in terms of total neutron fluence and ambient dose equivalent rates) and fluence-averaged energies.

  2. Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

    International Nuclear Information System (INIS)

    Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

    2012-01-01

    Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

  3. Computational modeling of the axial-cylindrical inertial electrostatic confinement fusion neutron generator

    Science.gov (United States)

    Bromley, Blair Patrick

    2001-12-01

    The axial-cylindrical Inertial Electrostatic Confinement fusion neutron generator (IEC C-Device) is a high- voltage, low-pressure glow discharge device that produces neutrons from the deuterium-deuterium fusion reaction. Such a neutron source has potential applications for neutron activation analysis and capture therapies for cancer treatment. The IEC C-Device operating with deuterium fuel is modeled with the CHIMP computer code developed and written completely by the author to predict the fusion neutron generation rate and the plasma physics behavior using fundamental first principles. The CHIMP code is a time-dependent, spatially two-dimensional (r,z), particle-in-cell, Monte-Carlo-Collision (PIC-MCC) direct simulation model. The effects of secondary electron emission due to ion and electron impact on the metal electrodes and the glass walls and charge build-up on the glass wall are included. Either monatomic or molecular ions and electrons are modeled in a monatomic or molecular background neutral deuterium gas. CHIMP code predictions are compared against experimental results for the C-Device operating between 10 and 30 kV of anode voltage, between 10 and 40 mA of electrode current, and between 0.29 and 1.1 milliTorr of deuterium gas pressure. A calibration factor for the pressure accounts for the calibration of the ionization pressure gauge in the experiment, and an estimated pressure drop between the main chamber of the C-Device and the pressure gauge that is downstream of the exhaust port. Upgraded versions of the CHIMP code which have modifications to the algorithms for the boundary conditions, and which include charge exchange processes, and the contribution of fast neutrals to the neutron generation rate are also tested against several experimental data points. Although the CHIMP code gives predictions for the neutron generation rate that exhibit the same near-linear trends with current found in the experiment, it is apparent that at least five types of

  4. Proceedings of the 14. International Symposium on the Interaction of Fast Neutrons with Nuclei - Neutron Generators and Application - organized by the Technical University of Dresden

    International Nuclear Information System (INIS)

    Seeliger, D.; Jahn, U.

    1985-07-01

    The symposium was devoted to current problems of intense fast neutron sources, especially 14 MeV DT-neutron generators, and their broad spectrum of application in nuclear physics. 56 participants from 12 countries and the IAEA demonstrate the high interest on this selected topics. The submitted contributions can be divided into two general parts. The first one gives a review about the different possibilities of the technical and technological solution in development, the present status of operation and also the problems connected with the use of intense neutron sources. Various experimental arrangements for neutron spectroscopy, determination of nuclear data and theoretical aspects are the content of the second part. The participation in this meeting of designer and operators on the one hand and users of neutron sources on the other hand was a good choice and stimulated productive discussions during the conference. (author)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  6. Biophysical interpretation of the response of Chinese hamster cells to 24 keV neutrons

    International Nuclear Information System (INIS)

    Holt, P.D.

    1988-01-01

    The response of V79 Chinese hamster cells to a 24 keV neutron spectrum has been compared with data for the response of V79 cells to a range of higher neutron energies (up to 15 MeV). The linear energy transfer (LET) distributions of the neutron spectra were calculated and the expected responses of the cells to the different spectra were calculated using published track-segment data on the response of V79 cells to charged particles with various LET values. The response of the cells to 24 keV neutrons was predicted satisfactorily by the LET distribution, in spite of the fact that the maximum range of the recoil protons is only 0.5 μm. The response was not correctly predicted by the microdosimetric parameter y-bar D * evaluated in a 1 μm diameter sphere. (author)

  7. Procedure and apparatus for the examination of underground formations with a neutron generator

    International Nuclear Information System (INIS)

    Culver, R.B.

    1978-01-01

    A pulsed neutron generator in a well logging instrument, pulsed at a clock frequency of 20 KHz is described. Inelastic scatter gamma rays are detected during a first time interval coinciding with the neutron source being on and capture gamma rays are measured during a second interval subsequent to the end of each neutron burst. Only a single detected pulse, assuming detection occurs, is transmitted during each of the two detection intervals. Sync pulses are generated in the well logging instrument scaled down to a frequency of 200 Hz for transmission to the earth's surface. At the earth's surface, the scaled-down sync pulses are applied to a phase locked loop system for regenerating the sync pulses to the same frequency as that of the clock frequency used to pulse the neutron source and to open the detection gates in the borehole instrument. The regenerated sync pulses are used in the surface instrumentation to route the pulses occurring in the inelastic interval into one section of a multichannel analyzer memory and the pulses occurring in the capture interval into another section of the multichannel analyzer. The use of memory address decoders, subtractors and ratio circuits enables both a carbon/oxygen ratio and a silicon/calcium ratio to be struck substantially free of background radiation. (Auth.)

  8. Research on determine the absolute neutron output of distributed pulse generators

    International Nuclear Information System (INIS)

    Li Bojun; Tang Zhangkui; Wang Dong; Yang Gaozhao; Peng Taiping

    2009-01-01

    In order to determine the absolute neutron output of distributed pulse generators, we deduced equivalent length to deal with experimental data, according to the different layout and weighting of multiple pulse generators. The deposited energy in scintillation crystal and the integral flux which drilling through crystal interface was simulated by MCNP code. The result shows the simulated proportion of different distributed pulse generators is approximately agreed with experimental data. The validity of the equivalent length model was proved by the consistent results between calculation and experimental data. (authors)

  9. Generating transverse response explicitly from harmonic oscillators

    Science.gov (United States)

    Yao, Yuan; Tang, Ying; Ao, Ping

    2017-10-01

    We obtain stochastic dynamics from a system-plus-bath mechanism as an extension of the Caldeira-Leggett (CL) model in the classical regime. An effective magnetic field and response functions with both longitudinal and transverse parts are exactly generated from the bath of harmonic oscillators. The effective magnetic field and transverse response are antisymmetric matrices: the former is explicitly time-independent corresponding to the geometric magnetism, while the latter can have memory. The present model can be reduced to previous representative examples of stochastic dynamics describing nonequilibrium processes. Our results demonstrate that a system coupled with a bath of harmonic oscillators is a general approach to studying stochastic dynamics, and provides a method to experimentally implement an effective magnetic field from coupling to the environment.

  10. A continuously self regenerating high-flux neutron-generator facility

    Science.gov (United States)

    Rogers, A. M.; Becker, T. A.; Bernstein, L. A.; van Bibber, K.; Bleuel, D. L.; Chen, A. X.; Daub, B. H.; Goldblum, B. L.; Firestone, R. B.; Leung, K.-N.; Renne, P. R.; Waltz, C.

    2013-10-01

    A facility based on a next-generation, high-flux D-D neutron generator (HFNG) is being constructed at UC Berkeley. The current generator, designed around two RF-driven multicusp deuterium ion sources, is capable of producing a neutron output of >1011 n/s. A specially designed titanium-coated copper target located between the ion sources accelerates D+ ions up to 150 keV, generating 2.45 MeV neutrons through the d(d,3He)n fusion reaction. Deuterium in the target is self loaded and regenerating through ion implantation, enabling stable and continuous long-term operation. The proposed science program is focused on pioneering advances in the 40Ar/39Ar dating technique for geochronology, new nuclear data measurements, basic nuclear science research including statistical model studies of radiative-strength functions and level densities, and education. An overview of the facility and its unique capabilities as well as first measurements from the HFNG commissioning will be presented. Work supported by NSF Grant No. EAR-0960138, U.S. DOE LBL Contract No. DE-AC02-05CH11231, and U.S. DOE LLNL Contract No. DE-AC52-07NA27344.

  11. Dosimetric response evaluation of tooth enamel for accelerator-based neutron radiation

    International Nuclear Information System (INIS)

    Khan, R.F.H.; Rink, W.J.; Boreham, D.R.

    2003-01-01

    To study the neutron response of human tooth enamel, a number of experiments with an accelerator-based neutron source have been designed. The neutron beam was produced with the low gamma yield, 7 Li(p,n) 7 Be type thick target, using the 3 MV McMaster K.N. Van de Graaff accelerator. The dosimetry was done using a pre-calibrated snoopy type neutron dosimeter. Neutron irradiation induces a dosimetric signal in the tooth enamel at the same defect site as gamma produced damage with the same g-values (g parallel =1.9973, width 0.4 mT g perpendicular =2.002, width 0.3 mT). The dosimetric signal grows linearly with neutron dose from 6-35 Gy tissue dose. Dosimetric response in two different grain sizes (300-500 μm, and grains <4 mm) has shown increased dosimetric amplitude in the larger grains. Dose build up effect on tooth inside the mouth due to cheek was simulated by placing a 4 mm thick paraffin wax layer between the beam and tooth, but had little effect. These results show that for mean neutron energy of 280 keV, the relative neutron response of the human tooth enamel ranges from 8% to 12% of the equivalent gamma ray response

  12. Variable dead time counters. 1 - theoretical responses and the effects of neutron multiplication

    International Nuclear Information System (INIS)

    Lees, E.W.; Hooton, B.W.

    1978-10-01

    A theoretical expression is derived for calculating the response of any variable dead time counter (VDC) used in the passive assay of plutonium by neutron counting of the natural spontaneous fission activity. The effects of neutron multiplication in the sample arising from interactions of the original spontaneous fission neutrons is shown to modify the linear relationship between VDC signal and Pu mass. Numerical examples are shown for the Euratom VDC and a systematic investigation of the various factors affecting neutron multiplication is reported. Limited comparisons between the calculations and experimental data indicate provisional validity of the calculations. (author)

  13. Experimental research of plastic scintillation detector loaded 6Li neutron energy response

    International Nuclear Information System (INIS)

    Wang Lizong; Zhang Chuanfei; Peng Taiping; Guo Cun; Yang Hongqiong; Zhang Jianhua

    2005-01-01

    A new plastic scintillator, plastic scintillator loaded 6 Li, is brought forward and developed in this paper in order to increase low energy neutron sensitivity. Neutron sensitivity of several plastic scintillation detectors loaded 6 Li new developed in neutron energy range 0.2 MeV-5.0 MeV are calibrated by direct current at serial accelerator. Energy response curves of the detectors are obtained in this experiment. It is shown that this new plastic scintillation detector can increase low energy neutron sensitivity in experimental results. (authors)

  14. Calibration of the time response functions of a quenched plastic scintillator for neutron time of flight

    CERN Document Server

    Chen, J B; Peng, H S; Tang, C H; Zhang, B H; Ding, Y K; Chen, M; Chen, H S; Li, C G; Wen, T S; Yu, R Z

    2002-01-01

    The time response functions of an ultrafast quenched plastic scintillation detector used to measure neutron time of flight spectra were calibrated by utilizing cosmic rays and implosion neutrons from DT-filled capsules at the Shenguang II laser facility. These sources could be regarded as delta function pulses due to their much narrower time widths than those of the time response functions of the detection system. The results showed that the detector responses to DT neutrons and to cosmic rays were 1.18 and 0.96 ns FWHM, respectively.

  15. A Compact Self-Driven Liquid Lithium Loop for Industrial Neutron Generation

    Science.gov (United States)

    Stemmley, Steven; Szott, Matt; Kalathiparambil, Kishor; Ahn, Chisung; Jurczyk, Brian; Ruzic, David

    2017-10-01

    A compact, closed liquid lithium loop has been developed at the University of Illinois to test and utilize the Li-7(d,n) reaction. The liquid metal loop is housed in a stainless steel trench module with embedded heating and cooling. The system was designed to handle large heat and particle fluxes for use in neutron generators as well as fusion devices, solely operating via thermo-electric MHD. The objectives of this project are two-fold, 1) produce a high energy, MeV-level, neutron source and 2) provide a self-healing, low Z, low recycling plasma facing component. The flowing volume will keep a fresh, clean, lithium surface allowing Li-7(d,n) reactions to occur as well as deuterium adsorption in the fluid, increasing the overall neutron output. Expected yields of this system are 107 n/s for 13.5 MeV neutrons and 108 n/s for 2.45 MeV neutrons. Previous work has shown that using a tapered trench design prevents dry out and allows for an increase in velocity of the fluid at the particle strike point. For heat fluxes on the order of 10's MW/m2, COMSOL models have shown that high enough velocities ( 70 cm/s) are attainable to prevent significant lithium evaporation. Future work will be aimed at addressing wettability issues of lithium in the trenches, experimentally determine the velocities required to prevent dry out, and determine the neutron output of the system. The preliminary results and discussion will be presented. DOE SBIR project DE-SC0013861.

  16. Prompt-period measurement of the Annular Core Research Reactor prompt neutron generation time

    International Nuclear Information System (INIS)

    Coats, R.L.; Talley, D.G.; Trowbridge, F.R.

    1994-07-01

    The prompt neutron generation time for the Annular Core Research Reactor was experimentally determined using a prompt-period technique. The resultant value of 25.5 μs agreed well with the analytically determined value of 24 μs. The three different methods of reactivity insertion determination yielded ±5% agreement in the experimental values of the prompt neutron generation time. Discrepancies observed in reactivity insertion values determined by the three methods used (transient rod position, relative delayed critical control rod positions, and relative transient rod and control rod positions) were investigated to a limited extent. Rod-shadowing and low power fuel/coolant heat-up were addressed as possible causes of the discrepancies

  17. Activation analysis course experiments with a 14-MeV neutron generator

    International Nuclear Information System (INIS)

    Miller, D.A.; Miller, G.E.

    1976-01-01

    The use of a 14 MeV neutron generator system in the radiochemistry teaching program of the Chemistry Department of the University of California at Irvine is described. Several different types of experiment are outlined to indicate the broad applicability of such a system to an instructional program in Chemistry. The program has encompassed instruction of undergraduates, graduate students and a Summer Institute Workshop for College Professors

  18. Comparison of calculated and measured spectral response and intrinsic efficiency for a boron-loaded plastic neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Kamykowski, E.A. (Grumman Corporate Research Center, Bethpage, NY (United States))

    1992-07-15

    Boron-loaded scintillators offer the potential for neutron spectrometers with a simplified, peak-shaped response. The Monte Carlo code, MCNP, has been used to calculate the detector characteristics of a scintillator made of a boron-loaded plastic, BC454, for neutrons between 1 and 7 MeV. Comparisons with measurements are made of spectral response for neutron energies between 4 and 6 MeV and of intrinsic efficiencies for neutrons up to 7 MeV. In order to compare the calculated spectra with measured data, enhancements to MCNP were introduced to generate tallies of light output spectra for recoil events terminating in a final capture by {sup 10}B. The comparison of measured and calculated spectra shows agreement in response shape, full width at half maximum, and recoil energy deposition. Intrinsic efficiencies measured to 7 MeV are also in agreement with the MCNP calculations. These results validate the code predictions and affirm the value of MCNP as a useful tool for development of sensor concepts based on boron-loaded plastics. (orig.).

  19. A Bayesian method to estimate the neutron response matrix of a single crystal CVD diamond detector

    International Nuclear Information System (INIS)

    Reginatto, Marcel; Araque, Jorge Guerrero; Nolte, Ralf; Zbořil, Miroslav; Zimbal, Andreas; Gagnon-Moisan, Francis

    2015-01-01

    Detectors made from artificial chemical vapor deposition (CVD) single crystal diamond are very promising candidates for applications where high resolution neutron spectrometry in very high neutron fluxes is required, for example in fusion research. We propose a Bayesian method to estimate the neutron response function of the detector for a continuous range of neutron energies (in our case, 10 MeV ≤ E n ≤ 16 MeV) based on a few measurements with quasi-monoenergetic neutrons. This method is needed because a complete set of measurements is not available and the alternative approach of using responses based on Monte Carlo calculations is not feasible. Our approach uses Bayesian signal-background separation techniques and radial basis function interpolation methods. We present the analysis of data measured at the PTB accelerator facility PIAF. The method is quite general and it can be applied to other particle detectors with similar characteristics

  20. Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV

    CERN Document Server

    Martinkovic, J

    2005-01-01

    Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.

  1. Dose-equivalent response CR-39 track detector for personnel neutron dosimetry

    International Nuclear Information System (INIS)

    Oda, K.; Ito, M.; Yoneda, H.; Miyake, H.; Yamamoto, J.; Tsuruta, T.

    1991-01-01

    A dose-equivalent response detector based on CR-39 has been designed to be applied for personnel neutron dosimetry. The intrinsic detection efficiency of bare CR-39 was first evaluated from irradiation experiments with monoenergetic neutrons and theoretical calculations. In the second step, the radiator effect was investigated for the purpose of sensitization to fast neutrons. A two-layer radiator consisting of deuterized dotriacontane (C 32 D 66 ) and polyethylene (CH 2 ) was designed. Finally, we made the CR-39 detector sensitive to thermal neutrons by doping with orthocarbone (B 10 H 12 C 2 ), and also estimated the contribution of albedo neutrons. It was found that the new detector - boron-doped CR-39 with the two-layer radiator - would have a flat response with an error of about 70% in a wide energy region, ranging from thermal to 15 MeV. (orig.)

  2. Track-etch dosemeter response to neutrons up to 300 MeV

    International Nuclear Information System (INIS)

    Devine, R.T.; Walker, S.; Staples, P.; Duran, M.; Mundis, R.; Miller, J.

    1996-01-01

    Electro-chemical and chemical track-etch dosemeters were obtained from commercial suppliers and exposed to neutrons produced at the LANTF WNR white neutron source at 15 degree with no shielding and filtered by polyethylene blocks of 2.5, 5.1, 10.2, 20.3 and 40.6 cin thickness. The neutron spectrum was determined using calculations. Mean energies from 28 to 300 MeV were produced. Dose was calculated from the NCRP-38 flux-to-dose conversion. The results are compared with NTA film which was exposed in the same configuration. The response of track etch dosimeters was found to reach a minimum and then rise as the average neutron energy increased. The response of the NTA film increased as the neutron energy increased

  3. Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

    Science.gov (United States)

    Shan, Qing; Chu, Shengnan; Jia, Wenbao

    2015-11-01

    Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Rehm, C.

    2001-01-01

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

  5. Pulsed neutron generators based on plasma focus devices of low energy

    International Nuclear Information System (INIS)

    Silva, Patricio; Moreno, Jose; Soto, Leopoldo

    2003-01-01

    The plasma focus is a pulsed neutron source especially suited for applications because it reduces the danger of contamination of conventional isotopic radioactive sources. As first stage of a program to design a repetitive pulsed neutron generator for industrial applications we constructed two very small plasma focus operating at an energy level of the order of a) tens of joules (PF-50J, 160nF capacitor bank, 20-35 kV, 32-100J, ∼150ns first quarter of period) and b) hundred of joules (PF-400J, 880nF, 20-35kV, 176-539J, ∼300ns first quarter of period). In this article we present results related to design and construction of these small plasma foci (PF-50J and PF-400J). Neutron yield vs. deuterium. pressure has been obtained, a maximum emission of the order of 7x10 4 and 10 6 neutrons per shot has been measured in the PF-50J and PF-400J respectively (author)

  6. Tritium generation and neutron measurements in Pd-Si under high deuterium gas pressure

    International Nuclear Information System (INIS)

    Claytor, T.N.; Tuggle, D.G.; Menlove, H.O.

    1991-01-01

    This paper summarizes some of the methods applicable for low level tritium detection needed in the search for anomalous fusion in metal hydrides. It is also intended to further detail our tritium and neutron results that have been obtained with the Pd-Si-D system, originally presented at earlier workshops. A measure of reproducibility that was not evident in our previous work has been achieved partially due to the better detection sensitivity afforded by the use of low tritium deuterium and partially from the fact that the foil-wafer cells can be made with nearly identical electrical characteristics. This reproducibility has allowed us to narrow the optimum conditions for the experiment. While this experiment is rather different from the ''standard'' electrolytic cell or the Ti gas hydride experiment, similarities exist in that non equilibrium conditions are sought and the tritium generation levels are low and neutron emission is extremely weak. In contrast to many electrochemical cell experiments, the system used in these experiments is completely sealed during operation and uses no electrolyte. The major improvements to the experiment have been the use of vary low tritium deuterium for the hydriding and the replacement of the aluminum neutron counter tubes with ones of stainless steel. These changes have resulted in pronounced improvements to the detection systems since the background tritium level in the gas has been reduced by a factor of 300 and the neutron background has been decreased by a factor of 14. 16 refs., 8 figs., 1 tab

  7. Use and generation of floor response spectra

    International Nuclear Information System (INIS)

    Ordonez Villalobos, A.

    1983-01-01

    One of the main objectives of the dynamic analysis of the structures of a nuclear power plant is the determination of the dynamic input that these structures transmit to the equipment and substructures they support, usually given as Floor Response Spectra (FRS). A close collaboration and feedback between the different groups that use and develop the FRS, is considered to be a very important factor in order to adapt the scope and content of the FRS to the precision required for a proper analysis or testing of the equipment; not only for the action of simple events but also for multiple combined actions. These aspects should be evaluated not only in the final stages of qualification of the equipment users schedules do not coincide with the schedules of the analysis group that develops the FRS. Different mechanisms of interchange of information and colaboration are suggested in order to optimize the availability, use and production of FRS. In the aspect of FRS generation, different procedures are reviewed including the direct procedures, not only for FRS but also for secondary FRS that are needed for the evaluation of equipment supported on other equipment or subsystems. It is concluded that in many cases, the direct procedures can be developed economically with the advantage that is easy to take into account the variability not only of the transfer function (including damping, stiffness and modal mass ratio). Different probabilities of excedence levels can be stabilized in order to obtain a more realistic dynamic response of the equipment. These last aspects can contribute to a more flexible procedure for the availability and generation of the FRS. (orig./HP)

  8. Evaluation of energy response of neutron rem monitor applied to high-energy accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakane, Yoshihiro; Harada, Yasunori; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    A neutron rem monitor was newly developed for applying to the high-intensity proton accelerator facility (J-PARC) that is under construction as a joint project between the Japan Atomic Energy Research Institute and the High Energy Accelerator Research Organization. To measure the dose rate accurately for wide energy range of neutrons from thermal to high-energy region, the neutron rem monitor was fabricated by adding a lead breeder layer to a conventional neutron rem monitor. The energy response of the monitor was evaluated by using neutron transport calculations for the energy range from thermal to 150 MeV. For verifying the results, the response was measured at neutron fields for the energy range from thermal to 65 MeV. The comparisons between the energy response and dose conversion coefficients show that the newly developed neutron rem monitor has a good performance in energy response up to 150 MeV, suggesting that the present study offered prospects of a practical fabrication of the rem monitor applicable to the high intensity proton accelerator facility. (author)

  9. Radiobiological response of fast neutrons on seedling growth of rice varieties with different amylose content

    International Nuclear Information System (INIS)

    Baradjanegara, A.A.; Sugiyanto, T.; Rahayu, S.

    1978-01-01

    Many studies are reported on radiation effects and on factors modifying the biological response of radiation in rice. However, little attention was directed towards studying effects of fast neutrons on seedling growth response of rice as a function of chemical constituents (e.g. amylose content). Experiments were conducted to investigate the dependency of amylose content in 4 rice cultivars on radiosensitivity to fast neutrons. From the results obtained a clear relationship between amylose content and sensitivity to fast neutrons could be shown. (author)

  10. In vitro studies of cellular response to DNA damage induced by boron neutron capture therapy

    International Nuclear Information System (INIS)

    Perona, M.; Pontiggia, O.; Carpano, M.; Thomasz, L.; Thorp, S.; Pozzi, E.; Simian, M.; Kahl, S.; Juvenal, G.; Pisarev, M.; Dagrosa, A.

    2011-01-01

    The aim of these studies was to evaluate the mechanisms of cellular response to DNA damage induced by BNCT. Thyroid carcinoma cells were incubated with 10 BPA or 10 BOPP and irradiated with thermal neutrons. The surviving fraction, the cell cycle distribution and the expression of p53 and Ku70 were analyzed. Different cellular responses were observed for each irradiated group. The decrease of Ku70 in the neutrons +BOPP group could play a role in the increase of sensitization to radiation.

  11. Model for Generation of Neutrons in a Compact Diode with Laser-Plasma Anode and Suppression of Electron Conduction Using a Permanent Cylindrical Magnet

    Science.gov (United States)

    Shikanov, A. E.; Vovchenko, E. D.; Kozlovskii, K. I.; Rashchikov, V. I.; Shatokhin, V. L.

    2018-04-01

    A model for acceleration of deuterons and generation of neutrons in a compact laser-plasma diode with electron isolation using magnetic field generated by a hollow cylindrical permanent magnet is presented. Experimental and computer-simulated neutron yields are compared for the diode structure under study. An accelerating neutron tube with a relatively high neutron generation efficiency can be constructed using suppression of electron conduction with the aid of a magnet placed in the vacuum volume.

  12. A novel fast-neutron tomography system based on a plastic scintillator array and a compact D–D neutron generator

    International Nuclear Information System (INIS)

    Adams, Robert; Zboray, Robert; Prasser, Horst-Michael

    2016-01-01

    Very few experimental imaging studies using a compact neutron generator have been published, and to the knowledge of the authors none have included tomography results using multiple projection angles. Radiography results with a neutron generator, scintillator screen, and camera can be seen in Bogolubov et al. (2005), Cremer et al. (2012), and Li et al. (2014). Comparable results with a position-sensitive photomultiplier tube can be seen in Popov et al. (2011). One study using an array of individual fast neutron detectors in the context of cargo scanning for security purposes is detailed in Eberhardt et al. (2005). In that case, however, the emphasis was on very large objects with a resolution on the order of 1 cm, whereas this study focuses on less massive objects and a finer spatial resolution. In Andersson et al. (2014) three fast neutron counters and a D–T generator were used to perform attenuation measurements of test phantoms. Based on the axisymmetry of the test phantoms, the single-projection information was used to calculate radial attenuation distributions of the object, which was compared with the known geometry. In this paper a fast-neutron tomography system based on an array of individual detectors and a purpose-designed compact D–D neutron generator is presented. Each of the 88 detectors consists of a plastic scintillator read out by two Silicon photomultipliers and a dedicated pulse-processing board. Data acquisition for all channels was handled by four single-board microcontrollers. Details of the individual detector design and testing are elaborated upon. Using the complete array, several fast-neutron images of test phantoms were reconstructed, one of which was compared with results using a Co-60 gamma source. The system was shown to be capable of 2 mm resolution, with exposure times on the order of several hours per reconstructed tomogram. Details about these measurements and the analysis of the reconstructed images are given, along with a

  13. Quantitative radiation dose-response relationships for normal tissues in man - I. Gustatory tissues response during photon and neutron radiotherapy

    International Nuclear Information System (INIS)

    Mossman, K.L.

    1982-01-01

    Quantitative radiation dose-response curves for normal gustatory tissue in man were studied. Taste function, expressed as taste loss, was evaluated in 84 patients who were given either photon or neutron radiotherapy for tumors in the head and neck region. Patients were treated to average tumor doses of 6600 cGy (photon) or 2200 cGy intervals for photon patients and 320-cGy intervals for neutron patients during radiotherapy. The dose-response curves for photons and neutrons were analyzed by fitting a four-parameter logistic equation to the data. Photon and neutron curves differed principally in their relative position along the dose axis. Comparison of the dose-response curves were made by determination of RBE. At 320 cGy, the lowest neutron dose at which taste measurements were made, RBE = 5.7. If this RBE is correct, then the therapeutic gain factor may be equal to or less than 1, indicating no biological advantage in using neutrons over photons for this normal tissue. These studies suggest measurements of taste function and evaluation of dose-response relationships may also be useful in quantitatively evaluating the efficacy of chemical modifiers of radiation response such as hypoxic cell radiosensitizers and radioprotectors

  14. Novel design concepts for generating intense accelerator based beams of mono-energetic fast neutrons

    International Nuclear Information System (INIS)

    Franklyn, C.B.; Govender, K.; Guzek, J.; Beer, A. de; Tapper, U.A.S.

    2001-01-01

    are presented elsewhere. In the case of an RFQ accelerator, operating in a pulsed bunched mode, a suitable shutter mechanism can be used to effectively isolate the gas target between beam pulses and thus considerably reduce the gas load on a differentially pumped system whilst still maintaining the target at pressures up to ∼1.2 bar. Such a system operating on a 2% duty cycle RFQ system has been implemented. To go to even higher gas pressure or higher accelerator duty cycle, further improvements to the gas target system, in the form of a plasma window, have been investigated and are being implemented. The RFQ linear accelerator presently utilised delivers a maximum average beam current of 100 μA of 3.6-4.9 MeV deuterons, dependent on the phase coupling between the two accelerating cavities. In a 30 mm long deuterium gas cell, operating at a pressure of 1.2 bar, the expected neutron emission is ∼10 10 s 1 , into the full solid angle. A maximum neutron energy obtained in the reaction D(d,n) 3 He would be 8.1 MeV with the spread of ∼750 keV. kinematics, approximately 50% of the primary neutron beam is emitted into a 20 deg. forward cone. This translates to the expected neutron densities in excess of 10 7 n.s -1 cm -2 some 10-20 cm away from the gas cell. Beam quality is high, with the slow neutron and gamma-ray components below 10% of the total primary fast neutron beam. The fast neutron energy spread (and the total neutron output) can be tailored to a specific application by adjusting either the gas cell length or the target gas pressure. The robust design and reasonable cost make the described neutron source a very attractive choice for variety of applications, such as mineral identification, material diagnostics (complementing thermal neutron radiography), and isotope generation. These activities are currently being pursued at NECSA along with close collaboration with academic institutions and industry. (author)

  15. Generation of thermonuclear fusion neutrons by means of a pure explosion. Part 2. Experimental results

    International Nuclear Information System (INIS)

    Derentowicz, H.; Kaliski, S.; Wolski, J.; Ziolkowski, Z.

    1977-01-01

    This paper presents the experimental results of the generation of a thermonuclear fusion neutrons by means of explosion. The experimental set is based on a quasi-spherical experiment in which a polyethylene layer is shot into a conic region hollowed out in a golden target and filled with deuterium gas. The speeding-up system is based on shooting the conic liner onto the surface of the Cu cone in which the Mach wave is generated and propagates along the cone axis leading to an implosion velocity of the polyethylene layer of the order of (4 - 5).10 6 cm/s. This affords a 10 3 -multiple compression of the D 2 gas (p 0 approximately 1.2 atm) and a neutron emission of the order of 3.10 7 from a mass of about 10 -7 g. This result is in full agreement with theoretical estimates. This is the first published and documented experiment in which a neutron stream of thermonuclear fusion was obtained by means of a pure explosion. (author)

  16. Determination of tritium generation and release parameters at lithium CPS under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ponkratov, Yuriy, E-mail: ponkratov@nnc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Baklanov, Viktor; Skakov, Mazhyn; Kulsartov, Timur; Tazhibayeva, Irina; Gordienko, Yuriy; Zaurbekova, Zhanna; Tulubayev, Yevgeniy [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Chikhray, Yevgeniy [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan); Lyublinski, Igor [JSC “Star”, Moscow (Russian Federation); NRNU “MEPhI”, Moscow (Russian Federation); Vertkov, Alexey [JSC “Star”, Moscow (Russian Federation)

    2016-11-01

    Highlights: • The main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor is described in paper. • In the experiments a very small tritium release was fixed likely due to its high solubility in liquid lithium. • If the lithium CPS will be used as a plasma facing material in temperature range up to 773 K under neutron irradiation only helium will release from lithium CPS into a vacuum chamber. - Abstract: This paper describes the main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor. The experiments were carried out using the method of mass-spectrometric registration of released gases and using a specially constructed ampoule device. Irradiation was carried out at different reactor thermal powers (1, 2 and 6 MW) and sample temperatures from 473 to 773 K. In the experiments a very small tritium release was detected likely due to its high solubility in liquid lithium. It can be caused by formation of lithium tritide during tritium diffusion to the lithium surface.

  17. Neutron therapy

    International Nuclear Information System (INIS)

    Riesler, Rudi

    1995-01-01

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

  18. Assessment of NJOY generated neutron heating factors based on JEF/EFF-1

    International Nuclear Information System (INIS)

    Vontobel, P.

    1990-01-01

    Using the NJOY nuclear data processing system, a coupled neutron-photon multigroup MATXS-formatted nuclear data library was generated based on the files JEF/EFF-1. The neutron heating factors contained in this VITAMIN-J structured library are compared with those of MACLIB-IV. The main differences are due to the included decay heat of shortlived reaction products in MACKLIB-IV and/or due to too high/low photon production data of some JEF/EFF-1 isotopes. It is recommended to check carefully the energy balance of new evaluations containing photon production data. How this can be done with the help of the NJOY HEATR module is shown in an example. (author) 35 figs., 9 refs

  19. Novel methods for improvement of a Penning ion source for neutron generator applications.

    Science.gov (United States)

    Sy, A; Ji, Q; Persaud, A; Waldmann, O; Schenkel, T

    2012-02-01

    Penning ion source performance for neutron generator applications is characterized by the atomic ion fraction and beam current density, providing two paths by which source performance can be improved for increased neutron yields. We have fabricated a Penning ion source to investigate novel methods for improving source performance, including optimization of wall materials and electrode geometry, advanced magnetic confinement, and integration of field emitter arrays for electron injection. Effects of several electrode geometries on discharge characteristics and extracted ion current were studied. Additional magnetic confinement resulted in a factor of two increase in beam current density. First results indicate unchanged proton fraction and increased beam current density due to electron injection from carbon nanofiber arrays.

  20. Optimize of Deuteron Current of 150 keV, 1 mA Neutron Generator

    International Nuclear Information System (INIS)

    Sri Sulamdari; Djasiman

    2003-01-01

    It has been characterized a 150 keV/1 mA Neutron Generator. It has been used some local components, except accelerator tube and vacuum system. To produce neutron, it has been used a deuterium gas bombarded into tritium target through reaction 3 H(d,n) 4 He. For preliminary experiment, we used the air as an ion source. The beam current of deuteron as a function of process parameters are presented in this paper. It's found that the optimum beam current of deuteron was 1000 μA, and this conditions was achieved at accelerations voltage 30 kV, extraction voltage 5 kV, guide voltage -11 kV and vacuum 10 -6 mbar. (author)

  1. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  2. RBE and clinical response in radiotherapy with neutron beams

    International Nuclear Information System (INIS)

    Ellis, F.

    1984-01-01

    Consideration of the clinical results reported, when a cyclotron produced neutron beam was used for treatments in the pelvis region, suggested that a constant RBE of 3 should not have been used for all neutron doses. Instead a variable RBE, which increased from approximately 3 to 8 (with decreasing dose), should have been used. Although some of these RBE values are much higher than 3, they have been observed in clinical practice. An ''equivalent photon'' isodose plan was produced by employing a variable RBE and, by taking a TDF limit of 86 for bowel, an isoeffect plan was produced. This shows that in the clinical situation under consideration much of the pelvis was overdosed. Doses to tumour cells and late effects are also briefly considered. It is suggested that, in neutron therapy, both an ''equivalent photon'' isodose plan and an isoeffect plan should be produced prior to treatment. (author)

  3. NeuLAND prototype: response to fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jaehrling, Simon; Scheit, Heiko [Technische Universitaet Darmstadt (Germany); Aumann, Thomas [Technische Universitaet Darmstadt (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Boretzky, Konstanze; Heil, Michael; Kresan, Dmytro; Simon, Haik [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Gasparic, Igor [Technische Universitaet Darmstadt (Germany); Rudjer Boskovic Institute, Zagreb (Croatia); Collaboration: R3B-Collaboration

    2014-07-01

    Within the R3B collaboration (Reactions with Relativistic Radioactive Beams), a new neutron detector NeuLAND (New Large Area Neutron Detector) is being developed. It will be a fully active scintillation detector consisting of 3000 scintillator bars, arranged in 30 double layers. Within a double layer 50 bars are horizontal and 50 vertical orientated. The whole detector measures 2.5 x 2.5 x 3 m{sup 3}. A prototype with 150 NeuLAND bars was tested at GSI using quasi-mono-energetic neutrons with different energies from 200 to 1500 MeV stemming from quasi-free deuteron breakup reactions on a CH{sub 2} target. The experimental setup is described, and preliminary results for the time resolution and efficiency of the NeuLAND prototype detector are presented.

  4. Improvement of Sodium Neutronic Nuclear Data for the Computation of Generation IV Reactors

    International Nuclear Information System (INIS)

    Archier, P.

    2011-01-01

    The safety criteria to be met for Generation IV sodium fast reactors (SFR) require reduced and mastered uncertainties on neutronic quantities of interest. Part of these uncertainties come from nuclear data and, in the particular case of SFR, from sodium nuclear data, which show significant differences between available international libraries (JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0). The objective of this work is to improve the knowledge on sodium nuclear data for a better calculation of SFR neutronic parameters and reliable associated uncertainties. After an overview of existing 23 Na data, the impact of the differences is quantified, particularly on sodium void reactivity effects, with both deterministic and stochastic neutronic codes. Results show that it is necessary to completely re-evaluate sodium nuclear data. Several developments have been made in the evaluation code Conrad, to integrate new nuclear reactions models and their associated parameters and to perform adjustments with integral measurements. Following these developments, the analysis of differential data and the experimental uncertainties propagation have been performed with Conrad. The resolved resonances range has been extended up to 2 MeV and the continuum range begins directly beyond this energy. A new 23 Na evaluation and the associated multigroup covariances matrices were generated for future uncertainties calculations. The last part of this work focuses on the sodium void integral data feedback, using methods of integral data assimilation to reduce the uncertainties on sodium cross sections. This work ends with uncertainty calculations for industrial-like SFR, which show an improved prediction of their neutronic parameters with the new evaluation. (author) [fr

  5. Beyond Californium-A Neutron Generator Alternative for Dosimetry and Instrument Calibration in the U.S.

    Science.gov (United States)

    Piper, Roman K; Mozhayev, Andrey V; Murphy, Mark K; Thompson, Alan K

    2017-09-01

    Evaluations of neutron survey instruments, area monitors, and personal dosimeters rely on reference neutron radiations, which have evolved from the heavy reliance on (α,n) sources to a shared reliance on (α,n) and the spontaneous fission neutrons of californium-252 (Cf). Capable of producing high dose equivalent rates from an almost point source geometry, the characteristics of Cf are generally more favorable when compared to the use of (α,n) and (γ,n) sources or reactor-produced reference neutron radiations. Californium-252 is typically used in two standardized configurations: unmoderated, to yield a fission energy spectrum; or with the capsule placed within a heavy-water moderating sphere to produce a softened spectrum that is generally considered more appropriate for evaluating devices used in nuclear power plant work environments. The U.S. Department of Energy Cf Loan/Lease Program, a longtime origin of affordable Cf sources for research, testing and calibration, was terminated in 2009. Since then, high-activity sources have become increasingly cost-prohibitive for laboratories that formerly benefited from that program. Neutron generators, based on the D-T and D-D fusion reactions, have become economically competitive with Cf and are recognized internationally as important calibration and test standards. Researchers from the National Institute of Standards and Technology and the Pacific Northwest National Laboratory are jointly considering the practicality and technical challenges of implementing neutron generators as calibration standards in the U.S. This article reviews the characteristics of isotope-based neutron sources, possible isotope alternatives to Cf, and the rationale behind the increasing favor of electronically generated neutron options. The evaluation of a D-T system at PNNL has revealed characteristics that must be considered in adapting generators to the task of calibration and testing where accurate determination of a dosimetric quantity is

  6. Non-induction of radioadaptive response in zebrafish embryos by neutrons

    International Nuclear Information System (INIS)

    Ng, Candy Y.P.; Kong, Eva Y.; Kobayashi, Alisa; Suya, Noriyoshi; Uchihori, Yukio; Cheng, Shuk Han; Konishi, Teruaki; Yu, Kwan Ngok

    2016-01-01

    In vivo neutron-induced radioadaptive response (RAR) was studied using zebrafish (Danio rerio) embryos. The Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility at the National Institute of Radiological Sciences (NIRS), Japan, was employed to provide 2-MeV neutrons. Neutron doses of 0.6, 1, 25, 50 and 100 mGy were chosen as priming doses. An X-ray dose of 2 Gy was chosen as the challenging dose. Zebrafish embryos were dechorionated at 4 h post fertilization (hpf), irradiated with a chosen neutron dose at 5 hpf and the X-ray dose at 10 hpf. The responses of embryos were assessed at 25 hpf through the number of apoptotic signals. None of the neutron doses studied could induce RAR. Non-induction of RAR in embryos having received 0.6- and 1-mGy neutron doses was attributed to neutron-induced hormesis, which maintained the number of damaged cells at below the threshold for RAR induction. On the other hand, non-induction of RAR in embryos having received 25-, 50- and 100-mGy neutron doses was explained by gamma-ray hormesis, which mitigated neutron-induced damages through triggering high-fidelity DNA repair and removal of aberrant cells through apoptosis. Separate experimental results were obtained to verify that high-energy photons could disable RAR. Specifically, 5- or 10-mGy X-rays disabled the RAR induced by a priming dose of 0.88 mGy of alpha particles delivered to 5-hpf zebrafish embryos against a challenging dose of 2 Gy of X-rays delivered to the embryos at 10 hpf

  7. Determination of the spatial response of neutron based analysers using a Monte Carlo based method

    International Nuclear Information System (INIS)

    Tickner, James

    2000-01-01

    One of the principal advantages of using thermal neutron capture (TNC, also called prompt gamma neutron activation analysis or PGNAA) or neutron inelastic scattering (NIS) techniques for measuring elemental composition is the high penetrating power of both the incident neutrons and the resultant gamma-rays, which means that large sample volumes can be interrogated. Gauges based on these techniques are widely used in the mineral industry for on-line determination of the composition of bulk samples. However, attenuation of both neutrons and gamma-rays in the sample and geometric (source/detector distance) effects typically result in certain parts of the sample contributing more to the measured composition than others. In turn, this introduces errors in the determination of the composition of inhomogeneous samples. This paper discusses a combined Monte Carlo/analytical method for estimating the spatial response of a neutron gauge. Neutron propagation is handled using a Monte Carlo technique which allows an arbitrarily complex neutron source and gauge geometry to be specified. Gamma-ray production and detection is calculated analytically which leads to a dramatic increase in the efficiency of the method. As an example, the method is used to study ways of reducing the spatial sensitivity of on-belt composition measurements of cement raw meal

  8. Cadmium-Zinc-Telluride photon detector for epithermal neutron spectroscopy--pulse height response characterisation

    International Nuclear Information System (INIS)

    Tardocchi, M.; Pietropaolo, A.; Andreani, C.; Bracco, A.; D'Angelo, A.; Gorini, G.; Imberti, S.; Senesi, R.; Rhodes, N.J.; Schooneveld, E.M.

    2004-01-01

    The Resonance Detector Spectrometer was recently revised for neutron spectroscopic studies in the eV energy region. In this technique one makes use of a photon detector to record the gamma emission from analyser foils used as neutron-gamma converters. The pulse-height response of a Cadmium-Zinc-Telluride photon detector to neutron capture emission from 238 U and 197 Au analyser foils was characterised in the neutron energy range 1-200 eV. The experiment was performed on the VESUVIO spectrometer at the ISIS neutron-pulsed source. A biparametric data acquisition, specifically developed for these measurements, allowed the simultaneous measurements of both the neutron time of flight and γ pulse-height spectra. Through the analysis of the γ pulse-height spectra the main components of the signal associated with resonant and non-resonant neutron absorption were identified. It was also shown that, in principle, energy discrimination can be used to improve the signal to background ratio of the neutron time-of-flight measurement

  9. Timing and position response of a block detector for fast neutron time-of-flight imaging

    Energy Technology Data Exchange (ETDEWEB)

    Laubach, M.A., E-mail: mlaubach@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Hayward, J.P., E-mail: jhayward@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Zhang, X., E-mail: xzhang39@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Cates, J.W., E-mail: jcates7@vols.utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2014-11-01

    Our research effort seeks to improve the spatial and timing performance of a block detector made of a pixilated plastic scintillator (EJ-200), first demonstrated as part of Oak Ridge National Laboratory's Advanced Portable Neutron Imaging System. Improvement of the position and time response is necessary to achieve better resolution and contrast in the images of shielded special nuclear material. Time-of-flight is used to differentiate between gamma and different sources of neutrons (e.g., transmission and fission neutrons). Factors limiting the timing and position performance of the neutron detector have been revealed through simulations and measurements. Simulations have suggested that the degradation in the ability to resolve pixels in the neutron detector is due to those interactions occurring near the light guide. The energy deposition within the neutron detector is shown to affect position performance and imaging efficiency. This examination details how energy cuts improve the position performance and degrade the imaging efficiency. Measurements have shown the neutron detector to have a timing resolution of σ=238 ps. The majority of this timing uncertainty is from the depth-of-interaction (DOI) of the neutron which is confirmed by simulations and analytical calculations.

  10. The fast neutron response of 7LiF thermoluminescent dosemeters

    International Nuclear Information System (INIS)

    Knipe, A.D.

    1990-02-01

    A series of experiments was performed in 1982 to determine the neutron energy response of the 7 LiF thermoluminescent dosemeter (TLD) employed in European zero power fast reactor gamma-ray energy deposition studies. Preliminary results of this work were included in a 1985 International experimental data-base of TLD neutron sensitivities and provided the most consistent set of data within the 7 LiF compilation. A more detailed interpretation of these data has improved the results still further, giving a mean relative neutron to gamma efficiency of 0.118±0.005. The main objective of this re-evaluation was to establish recommended neutron energy response values to replace a 1974 data set for the analysis of in-core fast reactor measurements. This was achieved by combining the mean experimental relative efficiency of 0.118 with calculated energy dependent kerma factors. The kerma factors for the TLD were based on US National Bureau of Standards values and a composition determined by chemical analysis. Adoption of the revised neutron energy response data set produces a small increase in the measured gamma-ray energy deposition of typically 2% relative to the 1974 data. However, more importantly, the detailed analysis of the experimental response data has significantly improved confidence in the neutron corrections applied to in-core TLD gamma-ray energy deposition measurements. (author)

  11. Prototype Neutron Energy Spectrometer

    International Nuclear Information System (INIS)

    Mitchell, Stephen; Mukhopadhyay, Sanjoy; Maurer, Richard; Wolff, Ronald

    2010-01-01

    The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production (ship effect), (a, n) reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

  12. Prototype Neutron Energy Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

    2010-06-16

    The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production [ship effect], [a, n] reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

  13. Monte carlo calculation of energy-dependent response of high-sensitive neutron monitor, HISENS

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji; Ebisawa, Tohru; Kobayashi, Keiji; Koide, Hiroaki; Seo, Takeshi; Kawano, Shinji

    1988-01-01

    A highly sensitive neutron monitor system, HISENS, has been developed to measure leakage neutrons from nuclear facilities. The counter system of HISENS contains a detector bank which consists of ten cylindrical proportional counters filled with 10 atm 3 He gas and a paraffin moderator mounted in an aluminum case. The size of the detector bank is 56 cm high, 66 cm wide and 10 cm thick. It is revealed by a calibration experiment using an 241 Am-Be neutron source that the sensitivity of HISENS is about 2000 times as large as that of a typical commercial rem-counter. Since HISENS is designed to have a high sensitivity in a wide range of neutron energy, the shape of its energy dependent response curve cannot be matched to that of the dose equivalent conversion factor. To estimate dose equivalent values from neutron counts by HISENS, it is necessary to know the energy and angular characteristics of both HISENS and the neutron field. The area of one side of the detector bank is 3700 cm 2 and the detection efficiency in the constant region of the response curve is about 30 %. Thus, the sensitivity of HISENS for this energy range is 740 cps/(n/cm 2 /sec). This value indicates the extremely high sensitivity of HISENS as compared with exsisting highly sensitive neutron monitors. (Nogami, K.)

  14. A feasibility study of a deuterium-deuterium neutron generator-based boron neutron capture therapy system for treatment of brain tumors.

    Science.gov (United States)

    Hsieh, Mindy; Liu, Yingzi; Mostafaei, Farshad; Poulson, Jean M; Nie, Linda H

    2017-02-01

    Boron neutron capture therapy (BNCT) is a binary treatment modality that uses high LET particles to achieve tumor cell killing. Deuterium-deuterium (DD) compact neutron generators have advantages over nuclear reactors and large accelerators as the BNCT neutron source, such as their compact size, low cost, and relatively easy installation. The purpose of this study is to design a beam shaping assembly (BSA) for a DD neutron generator and assess the potential of a DD-based BNCT system using Monte Carlo (MC) simulations. The MC model consisted of a head phantom, a DD neutron source, and a BSA. The head phantom had tally cylinders along the centerline for computing neutron and photon fluences and calculating the dose as a function of depth. The head phantom was placed at 4 cm from the BSA. The neutron source was modeled to resemble the source of our current DD neutron generator. A BSA was designed to moderate and shape the 2.45-MeV DD neutrons to the epithermal (0.5 eV to 10 keV) range. The BSA had multiple components, including moderator, reflector, collimator, and filter. Various materials and configurations were tested for each component. Each BSA layout was assessed in terms of the in-air and in-phantom parameters. The maximum brain dose was limited to 12.5 Gray-Equivalent (Gy-Eq) and the skin dose to 18 Gy-Eq. The optimized BSA configuration included 30 cm of lead for reflector, 45 cm of LiF, and 10 cm of MgF 2 for moderator, 10 cm of lead for collimator, and 0.1 mm of cadmium for thermal neutron filter. Epithermal flux at the beam aperture was 1.0 × 10 5  n epi /cm 2 -s; thermal-to-epithermal neutron ratio was 0.05; fast neutron dose per epithermal was 5.5 × 10 -13  Gy-cm 2 /φ epi , and photon dose per epithermal was 2.4 × 10 -13  Gy-cm 2 /φ epi . The AD, AR, and the advantage depth dose rate were 12.1 cm, 3.7, and 3.2 × 10 -3  cGy-Eq/min, respectively. The maximum skin dose was 0.56 Gy-Eq. The DD neutron yield that is needed to

  15. Standard Test Method for Measuring Neutron Fluence and Average Energy from 3H(d,n)4He Neutron Generators by Radioactivation Techniques 1

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers a general procedure for the measurement of the fast-neutron fluence rate produced by neutron generators utilizing the 3H(d,n)4He reaction. Neutrons so produced are usually referred to as 14-MeV neutrons, but range in energy depending on a number of factors. This test method does not adequately cover fusion sources where the velocity of the plasma may be an important consideration. 1.2 This test method uses threshold activation reactions to determine the average energy of the neutrons and the neutron fluence at that energy. At least three activities, chosen from an appropriate set of dosimetry reactions, are required to characterize the average energy and fluence. The required activities are typically measured by gamma ray spectroscopy. 1.3 The measurement of reaction products in their metastable states is not covered. If the metastable state decays to the ground state, the ground state reaction may be used. 1.4 The values stated in SI units are to be regarded as standard. No oth...

  16. D-T neutron generator development for cancer therapy. 1980 annual progress report

    International Nuclear Information System (INIS)

    Bacon, F.M.; Walko, R.J.; Bickes, R.W. Jr.; Cowgill, D.F.; Riedel, A.A.; O'Hagan, J.B.

    1980-05-01

    This report summarizes the work completed during the first year of a two-year grant by NCI/HEW to investigate the feasibility of developing a D-T neutron generator for use in cancer therapy. Experiments have continued on the Target Test Facility (TTF) developed during a previous grant to investigate high-temperature metal hydrides for use as target materials. The high voltage reliability of the TTF has been improved so that 200 kV, 200 mA operation is now routine. In recent target tests, the D-D neutron production rate was measured to be > 1 x 10 11 /s, a rate that corresponds to a D-T neutron production rate of > 1 x 10 13 /s - the desired rate for use in cancer therapy. Deuterium concentration depth profiles in the target, measured during intense ion beam bombardment, show that deuterium is depleted near the surface of the target due to impurities implanted by the ion beam. Recent modifications of the duopigatron ion source to reduce secondary electron damage to the electrodes also improved the ion source efficiency by about 40%. An ultra high vacuum version of the TTF is now being constructed to determine if improved vacuum conditions will reduce ion source impurities to a sufficiently low level that the deuterium near the surface of the target is not depleted. Testing will begin in June 1980

  17. Investigation of the response of improved self-powered neutron detectors

    International Nuclear Information System (INIS)

    Erk, S.

    1982-01-01

    The self-powered neutron detectors have been successfully employed for the most important parameters both for neutron flux and flux fluence determination. Their preference for such measurements due to their simplicity, convenience in use, rigidity, voluminal smallness and low price. However, self-powered neutron detectors depend on the type used, can only follow the neutron flux changes with a certain delay when they are compared to fission chambers which are thought to be the best detectors. In this thesis, a system has been proposed and considered carefully in order to speed up the response time, in another word, to correct the detector response to a level very near to fission chamber performance, a circuitry has been realized in the frame of principles so forth and applied to the experiments carried out in the TR-1 Reactor. Their positive results are presented. (author)

  18. Application of the generator coordinate method to neutron-rich Se and Ge isotopes

    Directory of Open Access Journals (Sweden)

    Higashiyama Koji

    2014-03-01

    Full Text Available The quantum-number projected generator coordinate method (GCM is applied to the neutron-rich Se and Ge isotopes, where the monopole and quadrupole pairing plus quadrupole-quadrupole interaction is employed as an effective interaction. The energy spectra obtained by the GCM are compared to both the shell model results and the experimental data. The GCM reproduces well the energy levels of high-spin states as well as the low-lying states. The structure of the low-lying collective states is analyzed through the GCM wave functions.

  19. Thermal analysis of titanium drive-in target for D-D neutron generation.

    Science.gov (United States)

    Jung, N S; Kim, I J; Kim, S J; Choi, H D

    2010-01-01

    Thermal analysis was performed for a titanium drive-in target of a D-D neutron generator. Computational fluid dynamics code CFX-5 was used in this study. To define the heat flux term for the thermal analysis, beam current profile was measured. Temperature of the target was calculated at some of the operating conditions. The cooling performance of the target was evaluated by means of the comparison of the calculated maximum target temperature and the critical temperature of titanium. Copyright 2009 Elsevier Ltd. All rights reserved.

  20. A 150 kV Isolation Transformer for a Neutron Generator

    International Nuclear Information System (INIS)

    Dechthummarong, C.; Pratumtip, P.; Thongleurm, C.; Vichaimongkol, P.; Charoennugul, R.; Vilaithong, T.

    1998-01-01

    The work aims at the design and construction of a 150 kV isolation transformer for a neutron generator. The transformer windings are designed to use cylindrical layers with circular enamel copper wires. The insulation of the dry type transformer uses the epoxy resin for encapsulated winding. This insulation is non-flammable under temperature 350 degree celsius and the breakdown voltage is 10-18 kV/mm. This insulation is suitable for insulating high voltage. The design of provides the temperature rise of winding not exceeding 65 degree celsius for protection of the cracking of epoxy resin due to the expansion of winding

  1. Campbelling-type theory of fission chamber signals generated by neutron chains in a multiplying medium

    International Nuclear Information System (INIS)

    Pál, L.; Pázsit, I.

    2015-01-01

    The signals of fission chambers are usually evaluated with the help of the co-called Campbelling techniques. These are based on the Campbell theorem, which states that if the primary incoming events, generating the detector pulses, are independent, then relationships exist between the moments of various orders of the signal in the current mode. This gives the possibility to determine the mean value of the intensity of the detection events, which is proportional to the static flux, from the higher moments of the detector current, which has certain advantages. However, the main application area of fission chambers is measurements in power reactors where, as is well known, the individual detection events are not independent, due to the branching character of the neutron chains (neutron multiplication). Therefore it is of interest to extend the Campbelling-type theory for the case of correlated neutron events. Such a theory could address two questions: partly, to investigate the bias when the traditional Campbell techniques are used for correlated incoming events; and partly, to see whether the correlation properties of the detection events, which carry information on the multiplying medium, could be extracted from the measurements. This paper is devoted to the investigation of these questions. The results show that there is a potential possibility to extract the same information from fission chamber signals in the current mode as with the Rossi- or Feynman-alpha methods, or from coincidence and multiplicity measurements, which so far have required detectors working in the pulse mode. It is also shown that application of the standard Campbelling techniques to neutron detection in multiplying systems does not lead to an error for estimating the stationary flux as long as the detector is calibrated in in situ measurements

  2. The generation, validation and testing of a coupled 219-group neutron 36-group gamma ray AMPX-II library

    International Nuclear Information System (INIS)

    Panini, G.C.; Siciliano, F.; Lioi, A.

    1987-01-01

    The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations

  3. Laser-energy scaling law for neutrons generated from nano particles Coulomb-exploded by intense femtosecond laser pulses

    International Nuclear Information System (INIS)

    Sakabe, Shuji; Hashida, Masaki

    2015-01-01

    To discuss the feasibility of compact neutron sources the yield of laser produced neutrons is scaled by the laser energy. High-energy ions are generated by Coulomb explosion of clusters through intense femtosecond laser-cluster interactions. The laser energy scaling law of the neutron yield is estimated using the laser intensity scaling law for the energy of ions emitted from clusters Coulomb-exploded by an intense laser pulse. The neutron yield for D (D, n) He shows the potential of compact neutron sources with modern laser technology, and the yield for p (Li, n) Be shows much higher than that for Li (p, n) Be with the assumption of 500 nm-class cluster Coulomb explosion. (author)

  4. Progress on using deuteron-deuteron fusion generated neutrons for 40Ar/39Ar sample irradiation

    Science.gov (United States)

    Rutte, Daniel; Renne, Paul R.; Becker, Tim; Waltz, Cory; Ayllon Unzueta, Mauricio; Zimmerman, Susan; Hidy, Alan; Finkel, Robert; Bauer, Joseph D.; Bernstein, Lee; van Bibber, Karl

    2017-04-01

    We present progress on the development and proof of concept of a deuteron-deuteron fusion based neutron generator for 40Ar/39Ar sample irradiation. Irradiation with deuteron-deuteron fusion neutrons is anticipated to reduce Ar recoil and Ar production from interfering reactions. This will allow dating of smaller grains and increase accuracy and precision of the method. The instrument currently achieves neutron fluxes of ˜9×107 cm-2s-1 as determined by irradiation of indium foils and use of the activation reaction 115In(n,n')115mIn. Multiple foils and simulations were used to determine flux gradients in the sample chamber. A first experiment quantifying the loss of 39Ar is underway and will likely be available at the time of the presentation of this abstract. In ancillary experiments via irradiation of K salts and subsequent mass spectrometric analysis we determined the cross-sections of the 39K(n,p)39Ar reaction at ˜2.8 MeV to be 160 ± 35 mb (1σ). This result is in good agreement with bracketing cross-section data of ˜96 mb at ˜2.45 MeV and ˜270 mb at ˜4 MeV [Johnson et al., 1967; Dixon and Aitken, 1961 and Bass et al. 1964]. Our data disfavor a much lower value of ˜45 mb at 2.59 MeV [Lindström & Neuer, 1958]. In another ancillary experiment the cross section for 39K(n,α)36Cl at ˜2.8 MeV was determined as 11.7 ± 0.5 mb (1σ), which is significant for 40Ar/39Ar geochronology due to subsequent decay to 36Ar as well as for the determination of production rates of cosmogenic 36Cl. Additional experiments resolving the cross section functions on 39K between 1.5 and 3.6 MeV are on their way using the LICORNE neutron source of the IPN Orsay tandem accelerator. Results will likely be available at the time of the presentation of this abstract. While the neutron generator is designed for fluxes of ˜109 cm-2s-1, arcing in the sample chamber currently limits the power—straightforwardly correlated to the neutron flux—the generator can safely be run at. Further

  5. Options for a next generation neutron source for neutron scattering based on the projected linac facility at JAERI

    International Nuclear Information System (INIS)

    Mezei, F.; Watanabe, Noboru; Niimura, Nobuo; Morii, Yukio; Aizawa, Kazuya; Suzuki, Jun-ichi.

    1997-03-01

    Japan Atomic Energy Research Institute (JAERI) has a project to construct a high intensity proton accelerator to promote wide basic science using neutrons and nuclear power technologies such as radioactive nuclide transmutation. One of the most important field for utilization of neutron beam is neutron scattering. The energy and the averaged current obtained by the proton accelerator are 1.5 GeV and 4-5.3 mA, respectively and these provide 6-8 MW power. The repetition frequency is 50-60 Hz. Evaluation of options for the use of accelerators for neutron production for neutron scattering research and investigation of the neutron research opportunities offered by sharing the superconducting linac planned at JAERI were discussed. There are two ways of the utilization of proton beams for neutron scattering experiment. One is for long pulse spallation source (LPSS) and the other is for short pulse spallation source (SPSS). Quantitative evaluation of instrument performance with LPSS and SPSS was examined in the intensive discussion, calculations, workshop on this topics with Prof. F. Mezei who stayed at JAERI from October 24 to November 6, 1996. A report of the collaborative workshop will be also published separately. (author)

  6. The D-D Neutron Generator as an Alternative to Am(Li) Isotopic Neutron Source in the Active Well Coincidence Counter

    Energy Technology Data Exchange (ETDEWEB)

    McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cleveland, Steven L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The 235U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques due to the limited penetration of the item by the characteristic 235U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Collar (UNCL) and 252Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., 252Cf or Am(Li)] to achieve better measurement accuracies than are possible using gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The 252Cf shuffler systems rely on the use of high-output 252Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every five to seven years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. In this work, we made use of Oak Ridge National Laboratory’s (ORNL’s) Large Volume Active Well Coincidence Counter (LV-AWCC) and a commercially available deuterium-deuterium (D-D) neutron generator to examine the potential of the D-D neutron generator as an alternative to the isotopic sources. We present the performance of the LV-AWCC with D-D generator for the assay of 235U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium (DU), low enriched uranium (LEU), and highly enriched uranium (HEU) items.

  7. Restart of the chemical preparation process for the fabrication of ZnO varistors for ferroelectric neutron generator power supplies

    International Nuclear Information System (INIS)

    Lockwood, Steven John

    2005-01-01

    To date, all varistors used in ferroelectric neutron generators have been supplied from a single, proprietary source, General Electric Corporate Research and Development (GE CR and D). To protect against the vulnerability of a single source, Sandia initiated a program in the early 1980's to develop a second source for this material. A chemical preparation process for making homogeneous, high purity ZnO-based varistor powder was generated, scaled to production quantities, and transferred to external suppliers. In 1992, the chem-prep varistor program was suspended when it appeared there was sufficient inventory of GE CR and D material to supply ferroelectric neutron generator production for many years. In 1999, neutron generator production schedules increased substantially, resulting in a predicted exhaustion of the existing supply of varistor material within five years. The chem-prep program was restarted in January, 2000. The goals of the program were to (1) duplicate the chem-prep powder synthesis process that had been qualified for WR production, (2) demonstrate sintered billets from the chem-prep powder met requirements, (3) develop a process for rod fabrication and demonstrate that all component specifications could be met, and (4) optimize the process from powder synthesis through component fabrication for full-scale production. The first three of these goals have been met and are discussed in this report. A facility for the fabrication of production quantities of chem-prep powder has been established. All batches since the restart have met compositional requirements, but differences in sintering behavior between the original process and the restarted process were noted. Investigation into the equipment, precipitant stoichiometry, and powder processing procedures were not able to resolve the discrepancies. It was determined that the restarted process, which incorporated Na doping for electrical stability (a process that was not introduced until the end of the

  8. Angular dependence of dose equivalent response of an albedo neutron dosimeter

    International Nuclear Information System (INIS)

    Torres, B.A.; Boswell, E.; Schwartz, R.B.

    1994-01-01

    The ANSI provides procedures for testing the performance of dosimetry services. Although neutron dose equivalent angular response studies are not now mandated, future standards may well require that such studies be performed. Current studies with an albedo dosimeter will yield information regarding the angular dependence of dose equivalent response for this type of personnel dosimeter. Preliminary data for bare 252 Cf fluences show a marked decrease in dosimeter reading with increasing angle. The response decreased by an approximate factor of four. For the horizontal orientation, the same response was noted from both positive and negative angles. However, for the vertical orientation, the response was unexplainably assymetric. We are also examining the response of the personnel badge in moderated 252 Cf fluences. Responses from the moderated and unmoderated 252 Cf fields and theoretical calculations of the neutron angular response will be compared. This information will assist in building a data base for future comparisons of neutron angular responses with other neutron albedo dosimeters and phantoms

  9. Response of a BGO detector to photon and neutron sources simulations and measurements

    CERN Document Server

    Vincke, H H; Fabjan, Christian Wolfgang; Otto, T

    2002-01-01

    In this paper Monte Carlo simulations (FLUKA) and measurements of the response of a BGO detector are reported. %For the measurements different radioactive sources were used to irradiate the BGO crystal. For the measurements three low-energy photon emitters $\\left({}^{60}\\rm{Co},\\right.$ ${}^{54}\\rm{Mn},$ $\\left. {}^{137}\\rm{Cs}\\right)$ were used to irradiate the BGO from various distances and angles. The neutron response was measured with an Am--Be neutron source. Simulations of the experimental irradiations were carried out. Our study can also be considered as a benchmark for FLUKA in terms of its reliability to predict the detector response of a BGO scintillator.

  10. MONTE CARLO CALCULATION OF THE ENERGY RESPONSE OF THE NARF HURST-TYPE FAST- NEUTRON DOSIMETER

    Energy Technology Data Exchange (ETDEWEB)

    De Vries, T. W.

    1963-06-15

    The response function for the fast-neutron dosimeter was calculated by the Monte Carlo technique (Code K-52) and compared with a calculation based on the Bragg-Gray principle. The energy deposition spectra so obtained show that the response spectra become softer with increased incident neutron energy ahove 3 Mev. The K-52 calculated total res nu onse is more nearly constant with energy than the BraggGray response. The former increases 70 percent from 1 Mev to 14 Mev while the latter increases 135 percent over this energy range. (auth)

  11. Influence of different moderator materials on characteristics of neutron fluxes generated under irradiation of lead target with proton beams

    International Nuclear Information System (INIS)

    Sosnin, A.N.; Polanski, A.; Petrochenkov, S.A.

    2002-01-01

    Neutron fields generated in extended heavy (Z ≥ 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (diam. 8 cm x 20 cm or diam. 8 cm x 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin

  12. Influence of Different Moderator Materials on Characteristics of Neutron Fluxes Generated under Irradiation of Lead Target with Proton Beams

    CERN Document Server

    Sosnin, A N; Polanski, A; Petrochenkov, S A; Golovatyuk, V M; Krivopustov, M I; Bamblevski, V P; Westmeier, W; Odoj, R; Brandt, R; Robotham, H; Hashemi-Nezhad, S R; Zamani-Valassiadou, M

    2002-01-01

    Neutron fields generated in extended heavy (Z\\geq 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (\\varnothing 8 cm\\times 20 cm or \\varnothing 8cm\\times 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared in the present work. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin.

  13. Testing of a Code for the Calculation of Spectra of Neutrons Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.

  14. Effects of thermal and fast neutrons and of ENU on generations M3 and M4 of Lens culinaris (medicus)

    International Nuclear Information System (INIS)

    Uhlik, J.; Urban, J.

    1976-01-01

    Plants in which the selection of the most fertile plants had not been made in the preceding generations showed a significantly lower emergence rate in the M3 and M4 generation after an ethyl nitroso urea (ENU) application, in comparison with material treated with neutrons. In the evaluation of the plants obtained after an exposure to the most effective doses in the induction of chlorophyll mutants, significant differences of the average values in relation to the control were found in the M3 generation in the number of seeds per plant after the application of both neutron radiations and ENU. In addition, after the application of thermal neutrons and ENU a significant difference was found in the average values of plant weight. A difference in the overall range of variability in relation to the control was found in plant weight after the application of neutrons and ENU, and in seed weight after the application of ENU and fast neutrons. The differences between the treated plants and controls in the M4 generation plants with fusarium disease were insignificant. The evaluation of the progenies exposed to various doses of the highest mutation effectiveness showed in the M3 generation significant differences (in relation to the control) in the mean values of plant height, seed weight, plant weight, seed proportion in plants, in the bottom-pod insertion level, and in the number of pods set. Despite a considerable attack by fusarium disease, the greatest number of plants having more seeds than 50 was selected in the M4 generation of the material exposed to the dose of 8 fast neutrons (0.95% of plants) while in the control the proportion of highly fertile plants was only 0.05%. The widest range of overall variability in the characteristics under study was found after irradiation with thermal neutrons. From this viewpoint they can be recommended for wide practical utilization. (author)

  15. A Permanent-Magnet Microwave Ion Source for a Compact High-Yield Neutron Generator

    International Nuclear Information System (INIS)

    Waldmann, Ole; Ludewigt, Bernhard

    2010-01-01

    We present recent work on the development of a microwave ion source that will be used in a high-yield compact neutron generator for active interrogation applications. The sealed tube generator will be capable of producing high neutron yields, 5 · 10 11 n/s for D-T and ∼ 1 · 10 10 n/s for D-D reactions, while remaining transportable. We constructed a microwave ion source (2.45 GHz) with permanent magnets to provide the magnetic field strength of 87.5 mT necessary for satisfying the electron cyclotron resonance (ECR) condition. Microwave ion sources can produce high extracted beam currents at the low gas pressures required for sealed tube operation and at lower power levels than previously used RF-driven ion sources. A 100 mA deuterium/tritium beam will be extracted through a large slit (60 · 6 mm 2 ) to spread the beam power over a larger target area. This paper describes the design of the permanent-magnet microwave ion source and discusses the impact of the magnetic field design on the source performance. The required equivalent proton beam current density of 40 mA/cm 2 was extracted at a moderate microwave power of 400 W with an optimized magnetic field.

  16. Analysis of fast neutron-generated mutants at the Arabidopsis thaliana HY4 locus

    International Nuclear Information System (INIS)

    Bruggemann, E.; Handwerger, K.; Essex, C.; Storz, G.

    1996-01-01

    Ionizing radiation is expected to produce mutants with deletions or other chromosomal rearrangements. These mutants are useful for a variety of purposes, such as creating null alleles and cloning genes whose existence is known only from their mutant phenotype; however, only a few mutations generated by ionizing radiation have been characterized at the molecular level in Arabidopsis thaliana. Twenty fast neutron-generated alleles of the Arabidopsis HY4 locus, which encodes a blue light receptor, CRY1, were isolated and characterized. Nine of the mutant alleles displayed normal genetic behavior. The other 11 mutant alleles were poorly transmitted through the male gametophyte and were lethal in homozygous plants. Southern blot analysis demonstrated that alleles of the first group generally contain small or moderate-sized deletions at HY4, while alleles of the second group contain large deletions at this locus. These results demonstrate that fast neutrons can produce a range of deletions at a single locus in Arabidopsis. Many of these deletions would be suitable for cloning by genomic subtraction or representational difference analysis. The results also suggest the presence of an essential locus adjacent to HY4. (author)

  17. Solution Synthesis and Processing of PZT Materials for Neutron Generator Applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, M.A.; Ewsuk, K.G.; Montoya, T.V.; Moore, R.H.; Sipola, D.L.; Tuttle, B.A.; Voigt, J.A.

    1998-12-01

    A new solution synthesis route has been developed for the preparation of lead-based ferroelectric materials (patent filed). The process produces controlled stoichiometry precursor powders by non-aqueous precipitation. For a given ferroelectric material to be prepared, a metal acetate/alkoxide solution containing constituent metal species in the appropriate ratio is mixed with an oxalic acid/n-propanol precipitant solution. An oxalate coprecipitate is instantly fonned upon mixing that quantitatively removes the metals from solution. Most of the process development was focused on the synthesis and processing of niobium-substituted lead zirconate titanate with a Zr-to-Ti ratio of 95:5 (PNZT 95/5) that has an application in neutron generator power supplies. The process was scaled to produce 1.6 kg of the PNZT 95/5 powder using either a sen-ii-batch or a continuous precipitation scheme. Several of the PNZT 95/5 powder lots were processed into ceramic slug form. The slugs in turn were processed into components and characterized. The physical properties and electrical performance (including explosive functional testing of the components met the requirements set for the neutron generator application. Also, it has been demonstrated that the process is highly reproducible with respect to the properties of the powders it produces and the properties of the ceramics prepared from its powders. The work described in this report was funded by Sandia's Laboratory Directed Research and Development Program.

  18. Thermal analysis of Ti drive-in target for D-D neutron generation

    International Nuclear Information System (INIS)

    Jung, N.S.; Kim, I.J.; Kim, S.J.; Choi, H.D.

    2008-01-01

    Full text: Thermal analysis was performed for a Ti drive-in target of a D-D neutron generator. Numerical calculation was the only feasible way to obtain the information of the target temperature, since it was very difficult to measure the target temperature during neutron generation due to high voltage being applied to the target. Computational fluid dynamics code CFX-5 was used in this study. In order to define the heat flux term for the thermal analysis, the current profile of the ion beam was measured. The one-dimensional, integrated current profile was measured by using a single slit and a Faraday cup. The measured current profile was transformed into the axially symmetric two-dimensional distribution function by using the Abel inversion, which had the two-dimensional Gaussian function shape. Temperature distribution in the target was calculated at the operating condition. The influence of operational parameters like the ion beam energy, current, coolant mass flow rate and coolant inlet temperature on the target temperature was investigated

  19. Development of a framework for the neutronics analysis system for next generation (Contract research)

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Yokoyama, Kenji

    2007-11-01

    In JAEA, authors has been promoting the development of an object-oriented next-generation neutronics analysis system in order to reflect the latest methods and models of reactor analysis to basic designs and operations of fast reactors in the efficient and effective way. A purpose of the developing system is to effectively realize requirements that has been difficult to manage in the conventional systems, such as change of analysis targets and change of analysis modeling levels. For the realization of the requirements, the authors adopted the two-layer model which consists of a control layer written in the Python as an object-oriented scripting language and a solver layer in the C++ as a system programming language. After having studied the principle on the two-layer model in the next-generation neutronics analysis system, the authors designed and implemented a library that enabled transparent transfer of data objects between the two layers. In each layer, appropriate numerical library was used for better performance. In the present library, a model proxy was implemented to exchange internal data that is represented in different ways in each layer. With this mechanism of the model proxy, it confirmed that data exchange between the layers can be performed easily and effectively. (author)

  20. Measurement of the Neutron Component in a Shower Generated in a Lead Target by Relativistic Nuclear Beam

    International Nuclear Information System (INIS)

    Chultehm, D.; Damdinsurehn, Ts.; D'yachenko, V.M.; Ehnkhzhin, L.; Lomova, L.A.; Perelygin, V.P.; Tolstov, K.D.

    1994-01-01

    The present paper describes a method of determining the total number of neutrons generated in an extended lead target by relativistic nuclei and protons. It is shown that 101±20 neutrons per proton are produced in the target with the volume of 50x50x80 cm 3 at 3.65 GeV energy of protons. 11 refs., 14 figs., 1 tab

  1. Neutron spectra and cross sections for ice and clathrate generated from the synthetic spectrum and synthetic model for molecular solids

    International Nuclear Information System (INIS)

    Petriw, S; Cantargi, F; Granada, R

    2006-01-01

    We present here a Synthetic Model for Molecular Solids, aimed at the description of the interaction of thermal neutrons with this kind of systems.Simple representations of the molecular dynamical modes are used, in order to produce a fair description of neutron scattering kernels and cross sections with a minimum set of input data. Using those spectra, we have generated thermal libraries for M C N P [es

  2. Development of a system for simultaneously generating triple extreme conditions for neutron scattering experiments

    Energy Technology Data Exchange (ETDEWEB)

    Ichimura, Shigeju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    We have developed new system available for controlling sample environment during the neutron scattering experiments. The system can simultaneously generate triple extreme conditions of low temperature, high magnetic field and high pressure. The system consists of : (1) a liquid-helium cryostat which enables the sample temperature range of 1.7 K to 200 K, (2) a superconducting magnet providing a vertical field up to 5 Tesla with antisymmetric split-coil geometry for polarized-beam experiments, and (3) a non-magnetic clamping high-pressure cell designed with the aim of generating hydrostatic pressure up to 2.5 Gpa. In the workshop, we will report the outline of the system and some results of performance tests using the system at JRR-3M of JAERI. (author)

  3. Neutron scattering facility for the calibration of the response to nuclear recoils

    International Nuclear Information System (INIS)

    Jochum, J.; Feilitzsch, F. von; Huber, M.; Jagemann, T.; Lachenmaier, T.; Lanfranchi, J.-C.; Potzel, W.; Ruedig, A.; Schnagl, J.; Stark, M.; Wulandari, H.; Chambon, B.; Drain, D.; Gascon, J.; Jesus, M. de; Martineau, O.; Simon, E.; Stern, M.

    2002-01-01

    A possibility to search for elementary particles as dark matter candidates is to detect elastic scattering with cryogenic detectors. For the interpretation of the data one has to determine the detector response to nuclear recoils, the so-called quenching factors. They can differ for the heat-, for the scintillation- and for the ionization-signal and can be measured by scattering of neutrons. The CRESST- and the EDELWEISS-collaborations have set up a neutron scattering facility for cryogenic detectors at the tandem-accelerator of the Munich 'Maier-Leibniz-Labor.' The scattering angle and the time-of-flight of the neutrons are measured by an array of liquid scintillator cells. The pulsed high energy (11 MeV) neutron beam is created by nuclear reaction of a 11 B on a H 2 -gas target. The set-up and the results of first tests are presented

  4. A multi-detector neutron spectrometer with nearly isotropic response for environmental and workplace monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Ros, J.M., E-mail: jm.gomezros@ciemat.e [CIEMAT, Av. Complutense 22, 28040 Madrid (Spain); Bedogni, R. [INFN-LNF Frascati National Laboratory-U.F. Fisica Sanitaria, via E. Fermi n. 40, 00044 Frascati (Italy); Moraleda, M.; Delgado, A.; Romero, A. [CIEMAT, Av. Complutense 22, 28040 Madrid (Spain); Esposito, A. [INFN-LNF Frascati National Laboratory-U.F. Fisica Sanitaria, via E. Fermi n. 40, 00044 Frascati (Italy)

    2010-01-21

    This communication describes an improved design for a neutron spectrometer consisting of {sup 6}Li thermoluminescent dosemeters located at selected positions within a single moderating polyethylene sphere. The spatial arrangement of the dosemeters has been designed using the MCNPX Monte Carlo code to calculate the response matrix for 56 log-equidistant energies from 10{sup -9} to 100 MeV, looking for a configuration that permits to obtain a nearly isotropic response for neutrons in the energy range from thermal to 20 MeV. The feasibility of the proposed spectrometer and the isotropy of its response have been evaluated by simulating exposures to different reference and workplace neutron fields. The FRUIT code has been used for unfolding purposes. The results of the simulations as well as the experimental tests confirm the suitability of the prototype for environmental and workplace monitoring applications.

  5. Inter-pulse high-resolution gamma-ray spectra using a 14 MeV pulsed neutron generator

    Science.gov (United States)

    Evans, L.G.; Trombka, J.I.; Jensen, D.H.; Stephenson, W.A.; Hoover, R.A.; Mikesell, J.L.; Tanner, A.B.; Senftle, F.E.

    1984-01-01

    A neutron generator pulsed at 100 s-1 was suspended in an artificial borehole containing a 7.7 metric ton mixture of sand, aragonite, magnetite, sulfur, and salt. Two Ge(HP) gamma-ray detectors were used: one in a borehole sonde, and one at the outside wall of the sample tank opposite the neutron generator target. Gamma-ray spectra were collected by the outside detector during each of 10 discrete time windows during the 10 ms period following the onset of gamma-ray build-up after each neutron burst. The sample was measured first when dry and then when saturated with water. In the dry sample, gamma rays due to inelastic neutron scattering, neutron capture, and decay were counted during the first (150 ??s) time window. Subsequently only capture and decay gamma rays were observed. In the wet sample, only neutron capture and decay gamma rays were observed. Neutron capture gamma rays dominated the spectrum during the period from 150 to 400 ??s after the neutron burst in both samples, but decreased with time much more rapidly in the wet sample. A signal-to-noise-ratio (S/N) analysis indicates that optimum conditions for neutron capture analysis occurred in the 350-800 ??s window. A poor S/N in the first 100-150 ??s is due to a large background continuum during the first time interval. Time gating can be used to enhance gamma-ray spectra, depending on the nuclides in the target material and the reactions needed to produce them, and should improve the sensitivity of in situ well logging. ?? 1984.

  6. A Combined Shielding Design for a Neutron Generator and a Linear Accelerator at Soreq NRC

    International Nuclear Information System (INIS)

    Epstein, L.

    2014-01-01

    A new radiography facility is designed at Soreq NRC. The facility will hold a neutron generator that produces 1.73·109 n/s with an energy of 14 MeV and a linear accelerator that accelerates electrons to an energy of 9 MeV. The two radiation sources will be installed in 2 separate laboratories that will be built in an existing building. Each laboratory will have its own machine and control room. The dose rates around the sources were calculated using the FLUKA Monte Carlo code(1,2). The annual doses were calculated in several regions around the generator and the accelerator laboratories in accordance with the occupancy in each area. The calculated annual doses were compared with the dose limits specified in the Safety at Work Regulations(3) and the IAEC Standard for Protection against Ionizing Radiation. The shielding was designed to comply with the following dose constraints: 0.3 mSv/y for members of the public and 2 mSv/y for radiation workers. Each radiation source is planned to produce radiation for a maximum of 500 hours per year. The dose rate in the direct beam of the accelerator is 30 Gy/min at 1 m from the source and it will be surrounded by a collimator with an opening of 30N-tilde horizontally and 2 mm vertically, 3 m from the radiation source. The leakage radiation dose will not be greater than 1.5 mGy/min (0.005% of the direct beam, according to the manufacturer). The leakage radiation will be produced isotropically. The neutron generator will be surrounded by a shielding made of a 10 cm iron cylinder (density 7.87 g/cm3), surrounded by 50 cm of borated polyethylene (atomic percent: H (13.8%), C (82.2%), B (4%), density: 0.92 g/cm3) and 5 cm of lead (density 11.35 g/cm3). The neutron generator shielding was not designed or required in the present shielding design but was considered in the shielding calculations

  7. A novel fast-neutron tomography system based on a plastic scintillator array and a compact D-D neutron generator.

    Science.gov (United States)

    Adams, Robert; Zboray, Robert; Prasser, Horst-Michael

    2016-01-01

    Very few experimental imaging studies using a compact neutron generator have been published, and to the knowledge of the authors none have included tomography results using multiple projection angles. Radiography results with a neutron generator, scintillator screen, and camera can be seen in Bogolubov et al. (2005), Cremer et al. (2012), and Li et al. (2014). Comparable results with a position-sensitive photomultiplier tube can be seen in Popov et al. (2011). One study using an array of individual fast neutron detectors in the context of cargo scanning for security purposes is detailed in Eberhardt et al. (2005). In that case, however, the emphasis was on very large objects with a resolution on the order of 1cm, whereas this study focuses on less massive objects and a finer spatial resolution. In Andersson et al. (2014) three fast neutron counters and a D-T generator were used to perform attenuation measurements of test phantoms. Based on the axisymmetry of the test phantoms, the single-projection information was used to calculate radial attenuation distributions of the object, which was compared with the known geometry. In this paper a fast-neutron tomography system based on an array of individual detectors and a purpose-designed compact D-D neutron generator is presented. Each of the 88 detectors consists of a plastic scintillator read out by two Silicon photomultipliers and a dedicated pulse-processing board. Data acquisition for all channels was handled by four single-board microcontrollers. Details of the individual detector design and testing are elaborated upon. Using the complete array, several fast-neutron images of test phantoms were reconstructed, one of which was compared with results using a Co-60 gamma source. The system was shown to be capable of 2mm resolution, with exposure times on the order of several hours per reconstructed tomogram. Details about these measurements and the analysis of the reconstructed images are given, along with a discussion

  8. Calculation and applications of the frequency dependent neutron detector response functions

    International Nuclear Information System (INIS)

    Van Dam, H.; Van Hagen, T.H.J.J. der; Hoogenboom, J.E.; Keijzer, J.

    1994-01-01

    The theoretical basis is presented for the evaluation of the frequency dependent function that enables to calculate the response of a neutron detector to parametric fluctuations ('noise') or oscillations in reactor core. This function describes the 'field view' of a detector and can be calculated with a static transport code under certain conditions which are discussed. Two applications are presented: the response of an ex-core detector to void fraction fluctuations in a BWR and of both in and ex-core detectors to a rotating neutron absorber near or inside a research reactor core. (authors). 7 refs., 4 figs

  9. Neutron dosimetry at commercial nuclear plants. Final report of Subtask B: dosimeter response

    Energy Technology Data Exchange (ETDEWEB)

    Cummings, F.M.; Endres, G.W.R.; Brackenbush, L.W.

    1983-03-01

    As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of commercial nuclear plants. In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors. The reactors were operating at full power during the irradiations. Measurements were also performed with electronic instruments, the tissue equivalent proportional counter (TEPC), and portable remmeters, SNOOPY, RASCAL and PNR-4.

  10. Neutron dosimetry at commercial nuclear plants. Final report of Subtask B: dosimeter response

    International Nuclear Information System (INIS)

    Cummings, F.M.; Endres, G.W.R.; Brackenbush, L.W.

    1983-03-01

    As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of commercial nuclear plants. In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors. The reactors were operating at full power during the irradiations. Measurements were also performed with electronic instruments, the tissue equivalent proportional counter (TEPC), and portable remmeters, SNOOPY, RASCAL and PNR-4

  11. Evaluation of slow shutdown system flux detectors in Point Lepreau Generating Station - I: dynamic response characterization

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, V.N.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Comeau, D. [New Brunswick Power Nuclear, Point Lepreau, New Brunswick (Canada); McKay, J.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Taylor, D. [New Brunswick Power Nuclear, Point Lepreau, New Brunswick (Canada)

    2009-07-01

    CANDU reactors are protected against reactor overpower by two independent shutdown systems: Shut Down System 1 and 2 (SDS1 and SDS2). At the Point Lepreau Generating Station (PLGS), the shutdown systems can be actuated by measurements of the neutron flux by Platinum-clad Inconel In-Core Flux Detectors (ICFDs). These detectors have a complex dynamic behaviour, characterized by 'prompt' and 'delayed' components with respect to immediate changes in the in-core neutron flux. The dynamic response components need to be determined accurately in order to evaluate the effectiveness of the detectors for actuating the shutdown systems. The amplitudes of the prompt and the delayed components of individual detectors were estimated over a period of several years by comparison of archived detector response data with the computed local neutron flux evolution for SDS1 and SDS2 reactor trips. This was achieved by custom-designed algorithms. The results of this analysis show that the dynamic response of the detectors changes with irradiation, with the SDS2 detectors having 'prompt' signal components that decreased significantly with irradiation. Some general conclusions about detector aging effects are also drawn. (author)

  12. Obtaining the neutron time-of-flight instrument response function for a single D-T neutron utilizing n-alpha coincidence from the d(t, α) n nuclear reaction

    Science.gov (United States)

    Styron, Jedediah; Ruiz, Carlos; Hahn, Kelly; Cooper, Gary; Chandler, Gordon; Jones, Brent; McWatters, Bruce; Smith, Jenny; Vaughan, Jeremy

    2017-10-01

    A measured neutron time-of-flight (nTOF) signal is a convolution of the neutron reaction history and the instrument response function (IRF). For this work, the IRF was obtained by measuring single, D-T neutron events by utilizing n-alpha coincidence. The d(t, α) n nuclear reaction was produced at Sandia National Laboratories' Ion Beam Laboratory using a 300-keV Cockroft-Walton generator to accelerate a 2- μA beam, of 175-keV D + ions, into a stationary, 2.6- μm, ErT2 target. Comparison of these results to those obtained using cosmic-rays and photons will be discussed. Sandia National Laboratories.

  13. MODELING THE RADIATION SHIELDING OF BORON NEUTRON CAPTURE THERAPY BASED ON 2.4 MEV D-D NEUTRON GENERATOR FACILITY

    Directory of Open Access Journals (Sweden)

    Muhammad Mu’Alim

    2018-01-01

    PEMODELAN PERISAI RADIASI PADA FASILITAS BORON NEUTRON CAPTURE THERAPY BERBASIS GENERATOR NEUTRON D-D 2,4 MeV. Telah dimodelkan perisai radiasi pada fasilitas Boron Neutron Capture Therapy (BNCT berbasis reaksi D-D pada Neutron Generator 2,4 MeV dengan Beam Shaping Assembly (BSA yang telah didesain sebelumnya. Pemodelan ini dilakukan untuk memperoleh suatu desain perisai radiasi untuk fasilitas BNCT berbasis generator neutron 2,4 MeV. Pemodelan dilakukan dengan cara memvariasikan bahan dan ketebalan perisasi radiasi. Bahan yang dipilih adalah beton barit, parafin, polietilen terborasi dan timbal. Perhitungan dilakukan menggunakan program MCNPX dengan tally F4 untuk menentukan laju dosis yang keluar dari perisai radiasi. Desain periasi radiasi dinyatakan optimal jika radiasi yang dihasilkan diluar perisai radiasi tidak melebihi Nilai Batas Dosis (NBD yang telah ditentukan oleh BAPETEN. Hasilnya, diperoleh suatu desain perisai radiasi menggunakan lapisan utama beton barit setebal 100 cm yang mengelilingi ruangan 100 cm x 100 cm x 166,4 cm dan polietilen terborasi 40 cm yang mengelilingi bahan beton barit. Kemudian ditambahkan beton barit 10 cm dan polietilen terborasi 10 cm untuk mengurangi radiasi primer yang lurus dari BSA setelah keluar dari lapisan utama. Laju dosis terbesar adalah 4,58 μSv·jam-1 pada sel 227 dan laju dosis rata-rata yang dihasilkan adalah sebesar 0,65 µSv·jam-1. Nilai laju dosis tersebut masih dibawah ambang batas NBD yang diperbolehkan oleh BAPETEN untuk pekerja radiasi. Kata kunci: Perisai radiasi, tally, laju dosis radiasi, BSA, BNCT

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-21

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

  15. Importance diagrams - a novel presentation of the response of a material to neutron irradiation

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1998-01-01

    Activation of fusion materials following neutron irradiation is of great technological importance, especially in the study of safety and environmental impacts. Currently, activation calculations are performed for a particular neutron spectrum, appropriate to a region in a particular fusion device, which makes it difficult to extract generic information. The present work gives details of a method to present the dominant nuclides for the radiological responses or an irradiated material in a fashion that is independent of the neutron spectrum and almost independent of the flux. The importance diagrams show regions in the decay time versus neutron energy space where a nuclide contributes >50% of the response. The importance diagrams for pure iron and SS316 are described, and it is noted that the shapes of the various regions vary very little with the total neutron flux. Variation of the diagrams with irradiation time occurs at short decay times in a systematic fashion. The use of the diagrams in a realistic spectrum relies on an expansion, which while not generally true, does hold approximately for many of the nuclides of interest. The diagrams are therefore a valuable summary of the universal, device-independent, response of the materials, and when combined with pathway information give a comprehensive description of activation for that material. (orig.)

  16. Development and experimental validation of a monte carlo modeling of the neutron emission from a d-t generator

    Energy Technology Data Exchange (ETDEWEB)

    Remetti, Romolo; Lepore, Luigi [Sapienza University of Rome, Dept. SBAI, Via Antonio Scarpa 14, 00161 Rome (Italy); Cherubini, Nadia [ENEA CRE Casaccia, Nuclear Material Characterization Laboratory and Nuclear Waste Management, Via Anguillarese 301, 00123 Rome (Italy)

    2017-01-11

    An extensive use of Monte Carlo simulations led to the identification of a Thermo Scientific MP320 neutron generator MCNPX input deck. Such input deck is currently utilized at ENEA Casaccia Research Center for optimizing all the techniques and applications involving the device, in particular for explosives and drugs detection by fast neutrons. The working model of the generator was obtained thanks to a detailed representation of the MP320 internal components, and to the potentialities offered by the MCNPX code. Validation of the model was obtained by comparing simulated results vs. manufacturer's data, and vs. experimental tests. The aim of this work is explaining all the steps that led to those results, suggesting a procedure that might be extended to different models of neutron generators.

  17. Development and experimental validation of a monte carlo modeling of the neutron emission from a d-t generator

    Science.gov (United States)

    Remetti, Romolo; Lepore, Luigi; Cherubini, Nadia

    2017-01-01

    An extensive use of Monte Carlo simulations led to the identification of a Thermo Scientific MP320 neutron generator MCNPX input deck. Such input deck is currently utilized at ENEA Casaccia Research Center for optimizing all the techniques and applications involving the device, in particular for explosives and drugs detection by fast neutrons. The working model of the generator was obtained thanks to a detailed representation of the MP320 internal components, and to the potentialities offered by the MCNPX code. Validation of the model was obtained by comparing simulated results vs. manufacturer's data, and vs. experimental tests. The aim of this work is explaining all the steps that led to those results, suggesting a procedure that might be extended to different models of neutron generators.

  18. Experimental research of neutron yield and spectrum from deuterium gas-puff z-pinch on the GIT-12 generator at current above 2 MA

    Science.gov (United States)

    Cherdizov, R. K.; Fursov, F. I.; Kokshenev, V. A.; Kurmaev, N. E.; Labetsky, A. Yu; Ratakhin, N. A.; Shishlov, A. V.; Cikhardt, J.; Cikhardtova, B.; Klir, D.; Kravarik, J.; Kubes, P.; Rezac, K.; Dudkin, G. N.; Garapatsky, A. A.; Padalko, V. N.; Varlachev, V. A.

    2017-05-01

    The Z-pinch experiments with deuterium gas-puff surrounded by an outer plasma shell were carried out on the GIT-12 generator (Tomsk, Russia) at currents of 2 MA. The plasma shell consisting of hydrogen and carbon ions was formed by 48 plasma guns. The deuterium gas-puff was created by a fast electromagnetic valve. This configuration provides an efficient mode of the neutron production in DD reaction, and the neutron yield reaches a value above 1012 neutrons per shot. Neutron diagnostics included scintillation TOF detectors for determination of the neutron energy spectrum, bubble detectors BD-PND, a silver activation detector, and several activation samples for determination of the neutron yield analysed by a Sodium Iodide (NaI) and a high-purity Germanium (HPGe) detectors. Using this neutron diagnostic complex, we measured the total neutron yield and amount of high-energy neutrons.

  19. Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

    Science.gov (United States)

    Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

    2017-04-01

    The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

  20. Pulsed neutron generators based on the sealed chambers of plasma focus design with D and DT fillings

    International Nuclear Information System (INIS)

    Yurkov, D I; Dulatov, A K; Lemeshko, B D; Golikov, A V; Andreev, D A; Mikhailov, Yu V; Prokuratov, I A; Selifanov, A N

    2015-01-01

    Development of neutron generators using plasma focus (PF) chambers is being conducted in the All-Russia Scientific Research Institute of Automatics (VNIIA) during more than 25 years. PF is a source of soft and hard x-rays and neutrons 2.5 MeV (D) or 14 MeV (DT). Pulses of x-rays and neutrons have a duration of about several tens of nanoseconds, which defines the scope of such generators—the study of ultrafast processes. VNIIA has developed a series of pulse neutron generators covering the range of outputs 10 7 –10 12 n/pulse with resources on the order of 10 3 –10 4 switches, depending on purposes. Generators have weights in the range of 30–700 kg, which allows referring them to the class of transportable generators. Generators include sealed PF chambers, whose manufacture was mastered by VNIIA vacuum tube production plant. A number of optimized PF chambers, designed for use in generators with a certain yield of neutrons has been developed. The use of gas generator based on gas absorber of hydrogen isotopes, enabled to increase the self-life and resource of PF chambers. Currently, the PF chambers withstand up to 1000 switches and have the safety of not less than 5 years. Using a generator with a gas heater, significantly increased security of PF chambers, because deuterium-tritium mixture is released only during work, other times it is in a bound state in the working element of the gas generator. (paper)

  1. ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    International Nuclear Information System (INIS)

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-01-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

  2. First demonstration of laser engagement of 1-Hz-injected flying pellets and neutron generation

    Science.gov (United States)

    Komeda, Osamu; Nishimura, Yasuhiko; Mori, Yoshitaka; Hanayama, Ryohei; Ishii, Katsuhiro; Nakayama, Suisei; Kitagawa, Yoneyoshi; Sekine, Takashi; Sato, Nakahiro; Kurita, Takashi; Kawashima, Toshiyuki; Kan, Hirofumi; Nakamura, Naoki; Kondo, Takuya; Fujine, Manabu; Azuma, Hirozumi; Motohiro, Tomoyoshi; Hioki, Tatsumi; Kakeno, Mitsutaka; Sunahara, Atsushi; Sentoku, Yasuhiko; Miura, Eisuke

    2013-01-01

    Pellet injection and repetitive laser illumination are key technologies for realizing inertial fusion energy. Numerous studies have been conducted on target suppliers, injectors, and tracking systems for flying pellet engagement. Here we for the first time demonstrate the pellet injection, counter laser beams' engagement and neutron generation. Deuterated polystyrene (CD) bead pellets, after free-falling for a distance of 18 cm at 1 Hz, are successfully engaged by two counter laser beams from a diode-pumped, ultra-intense laser HAMA. The laser energy, pulse duration, wavelength, and the intensity are 0.63 J per beam, 104 fs, and 811 nm, 4.7 × 1018 W/cm2, respectively. The irradiated pellets produce D(d,n)3He-reacted neutrons with a maximum yield of 9.5 × 104/4π sr/shot. Moreover, the laser is found out to bore a straight channel with 10 μm-diameter through the 1-mm-diameter beads. The results indicate potentially useful technologies and findings for the next step in realizing inertial fusion energy. PMID:24008696

  3. 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)

  4. ENERGY RESPONSE OF FLUORESCENT NUCLEAR TRACK DETECTORS OF VARIOUS COLORATIONS TO MONOENERGETIC NEUTRONS.

    Science.gov (United States)

    Fomenko, V; Moreno, B; Million, M; Harrison, J; Akselrod, M

    2017-10-25

    The neutron-energy dependence of the track-counting sensitivity of fluorescent nuclear track detectors (FNTDs) at two ranges of Mg doping, resulting in different crystal colorations, was investigated. The performance of FNTDs was studied with the following converters: Li-glass for thermal to intermediate-energy neutrons, polyethylene for fast neutrons, and polytetrafluoroethylene (Teflon™) for photon- and radon-background subtraction. The irradiations with monoenergetic neutrons were performed at the National Physics Laboratory (NPL), UK. The energy range was varied from 144 keV to 16.5 MeV in the personal dose equivalent range from 1 to 3 mSv. Monte Carlo simulations were performed to model the response of FNTDs to monoenergetic neutrons. A good agreement with the experimental data was observed suggesting the development of a basic model for future MC studies. Further work will focus on increasing FNTD sensitivity to low-energy neutrons and developing a faster imaging technique for scanning larger areas to improve counting statistics. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  5. Investigating the response of Micromegas detector to low-energy neutrons using Monte Carlo simulation

    Science.gov (United States)

    Khezripour, S.; Negarestani, A.; Rezaie, M. R.

    2017-08-01

    Micromegas detector has recently been used for high-energy neutron (HEN) detection, but the aim of this research is to investigate the response of the Micromegas detector to low-energy neutron (LEN). For this purpose, a Micromegas detector (with air, P10, BF3, 3He and Ar/BF3 mixture) was optimized for the detection of 60 keV neutrons using the MCNP (Monte Carlo N Particle) code. The simulation results show that the optimum thickness of the cathode is 1 mm and the optimum of microgrid location is 100 μm above the anode. The output current of this detector for Ar (3%) + BF3 (97%) mixture is greater than the other ones. This mixture is considered as the appropriate gas for the Micromegas neutron detector providing the output current for 60 keV neutrons at the level of 97.8 nA per neutron. Consecuently, this detector can be introduced as LEN detector.

  6. Preliminary neutronic design of spock reactor: A nuclear system for space power generation

    International Nuclear Information System (INIS)

    Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

    2007-01-01

    Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state. The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out. The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO 2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B 4 C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg. The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth and a power spatial distribution were also discussed. Through the definition of an ideal reference K e ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on

  7. Response of CMS avalanche photo-diodes to low energy neutrons

    Science.gov (United States)

    Brown, R. M.; Deiters, K.; Ingram, Q.; Renker, D.

    2012-12-01

    The response of the Avalanche Photo-diodes (APDs) installed in the CMS detector at the LHC to neutrons from 241AmBe and 252Cf sources is reported. Signals in size equivalent to those of up to 106 photo-electrons with the nominal APD gain are observed. Measurements with an APD with the protective epoxy coating removed and with the source placed behind the APD show that there is an important response due to recoil protons from neutron interactions with the hydrogen in the epoxy, in addition to signals from neutron interactions with the silicon of the diode. The effective gain of these signals is much smaller than the diode's nominal gain.

  8. Response matrix of a multisphere neutron spectrometer with an 3 He proportional counter

    International Nuclear Information System (INIS)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A.

    2005-01-01

    The response matrix of a Bonner sphere spectrometer was calculated by use of the MCNP code. As thermal neutron counter, the spectrometer has a 3.2 cm-diameter 3 He-filled proportional counter which is located at the center of a set of polyethylene spheres. The response was calculated for 0, 3, 5, 6, 8, 10, 12, and 16 inches-diameter polyethylene spheres for neutrons whose energy goes from 10 -9 to 20 MeV. The response matrix was compared with a set of responses measured with several monoenergetic neutron sources. In this comparison the calculated matrix agrees with the experimental results. The matrix was also compared with the response matrix calculated for the PTB C spectrometer. Even though that calculation was carried out using a detailed model to describe the proportional counter; both matrices do agree, but small differences are observed in the bare case because of the difference in the model used during calculations. Other differences are in some spheres for 14.8 and 20 MeV neutrons, probably due to the differences in the cross sections used during both calculations. (Author) 28 refs., 1 tab., 6 figs

  9. Computed neutron response of spherical moderator-detector systems for radiation protection monitoring

    International Nuclear Information System (INIS)

    Dhairyawan, M.P.

    1979-01-01

    Neutrons of energies below 500 keV are important from the point of view of radiation protection of personnel working around reactors. However, as no neutron sources are available at lower energies, no measured values of neutron energy response are available between thermal and 0.5 MeV (but for Sb-Be source at 24 keV). The response functions in this range are, therefore, arrived at theoretically. After giving a comprehensive review of the work done in the field of response of moderated neutron detectors, a Monte Carlo method developed for this purpose is described and used to calculate energy response functions of the two spherical moderator-detector systems, namely, one using a central BF 3 counter and the other using 6 LiI(Eu) scintillator of 0.490 dia crystal. The polythene sphere diameter ranged from 2'' to 12''. The results obtained follow the trend predicted by other calculations and experiments, but are a definite improvement over them, because the most recent data on cross sections and angular distribution are used and the opacity of the detector i.e. the presence and size of the detector within the moderator is taken into account in the present calculations. The reasons for the discrepancies in the present results and those obtained earlier by other methods are discussed. The response of the Leake counter arrived at by the present method agrees very well with experimental calibration. (M.G.B.)

  10. Ion tail formation and its effect on 14-MeV neutron generation in D-3He plasmas

    International Nuclear Information System (INIS)

    Matsuura, H.; Nakao, Y.; Kudo, K.

    1992-01-01

    This paper reports on the triton distribution function in D- 3 He plasmas which is distorted from a Maxwellian owing to the presence of a 1.01-MeV birth component. The deuteron-triton reaction rate (i.e., 14-MeV neutron generation rate) in the plasma should be smaller than the values evaluated by assuming a Maxwellian triton distribution. A local Fokker-Planck calculation shows that although the degree of the decrease in 14-MeV neutron generation strongly depends on the plasma conditions and also on the energy loss mechanism, it becomes appreciable in actual burning plasmas

  11. Experimental and numerical investigations of radiation characteristics of Russian portable/compact pulsed neutron generators: ING-031, ING-07, ING-06 and ING-10-20-120

    International Nuclear Information System (INIS)

    Chernikova, D.; Romodanov, V.L.; Belevitin, A.G.; Afanas'ev, V.V.; Sakharov, V.K.; Bogolubov, E.P.; Ryzhkov, V.I.; Khasaev, T.O.; Sladkov, A.A.; Bitulev, A.A.

    2014-01-01

    The present paper discusses results of full-scale experimental and numerical investigations of influence of construction materials of portable pulsed neutron generators ING-031, ING-07, ING-06 and ING-10-20-120 (VNIIA, Russia) to their radiation characteristics formed during and after an operation (shutdown period). In particular, it is shown that an original monoenergetic isotropic angular distribution of neutrons emitted by TiT target changes into the significantly anisotropic angular distribution with a broad energy spectrum stretching to the thermal region. Along with the low-energetic neutron part, a significant amount of photons appears during the operation of generators. In the pulse mode of operation of neutron generator, a presence of the construction materials leads to the “tailing” of the original neutron pulse and the appearance of an accompanying photon pulse at ∼3ns after the instant neutron pulse. In addition to that, reactions of neutron capture and inelastic scattering lead to the creation of radioactive nuclides, such as 58 Co, 62 Cu, 64 Cu and 18 F, which form the so-called activation radiation. Thus, the selection of a portable neutron generator for a particular type of application has to be done considering radiation characteristics of the generator itself. This paper will be of interest to users of neutron generators, providing them with valuable information about limitations of a specific generator and with recommendations for improving the design and performance of the generator as a whole

  12. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  13. 1013/second 14 MeV neutron generator for cancer therapy

    International Nuclear Information System (INIS)

    Walko, R.J.; Bacon, F.M.; Bickes, R.W. Jr.; Cowgill, D.F.; Boers, J.E.; Riedel, A.A.; O'Hagan, J.B.

    1980-01-01

    A high output D-T neutron generator is being developed at Sandia National Laboratories as a cancer therapy tool. Preliminary results have been obtained with a high vacuum deuterium accelerator consisting of a duopigatron ion source, single stage accelerator and a scandium occluder target. The ion source and accelerating system have been optimized to reliably focus a 200 mA, 200 kV deuterium ion beam onto the scandium target 1.5 m away with a beam spot size of 10 -3 m 2 . The scandium target degrades due to implantation of impurities from the unbaked vacuum system. A UHV accelerator is being fabricated which will permit a 450 0 C bakeout to alleviate the impurity problem. This system will also be compatible with the use of tritium, a necessary component of the therapy machine

  14. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  15. Schlieren diagnostics of the Los Alamos hypersonic gas target neutron generator

    International Nuclear Information System (INIS)

    Haasz, A.A.; Lever, J.H.

    1981-01-01

    The gasdynamic behaviour of a planar model of the Los Alamos geometry hypersonic gas target neutron generator (GTNG) was investigated using Schlieren flow visualization photographs, static and total pressure and spill flow measurements. The model consisted of two symmetrical expansion nozzles with 220 μm throats producing a combined flow of about Mach 4 in the GTNG channel. Stagnation pressures of 100-800 kPa were used. Two basic flow configurations, spill line closed and spill line open, were studied in order to gain insight into the complex boundary layer development near the nozzle exit planes. Both flow configurations are discussed qualitatively, making use of the pressure measurements and theoretical analysis. (orig.)

  16. Project and construction of a pneumatic system for the transference of samples to a neutron generator

    International Nuclear Information System (INIS)

    Carvalho, A.N. de

    1983-01-01

    A prototype of a system for the transport of irradiated samples to and from a neutron generator, was constructed, using compressed air as propeller agent. Compressed air was injected through electrically driven values. The sample, transported by the pressure wave, was inserted into a PVC tube 50m long and weighing 23.0 g. The first tests were carried out in order to determine the times needed to transport the above-mentioned PVC support along a PVC tube of 3m length and 3/4 diameter for different air pressures applied; it was verified that for pressures between 3.0 and 8.0 kgf/cm 2 , transport times were always smaller than 2 seconds. These results showed the viability of constructing a definitive system, already projected. (C.L.B.) [pt

  17. Use of Aria to simulate laser weld pool dynamics for neutron generator production.

    Energy Technology Data Exchange (ETDEWEB)

    Noble, David R.; Notz, Patrick K.; Martinez, Mario J.; Kraynik, Andrew Michael

    2007-09-01

    This report documents the results for the FY07 ASC Integrated Codes Level 2 Milestone number 2354. The description for this milestone is, 'Demonstrate level set free surface tracking capabilities in ARIA to simulate the dynamics of the formation and time evolution of a weld pool in laser welding applications for neutron generator production'. The specialized boundary conditions and material properties for the laser welding application were implemented and verified by comparison with existing, two-dimensional applications. Analyses of stationary spot welds and traveling line welds were performed and the accuracy of the three-dimensional (3D) level set algorithm is assessed by comparison with 3D moving mesh calculations.

  18. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  19. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Frescura, G.M.; Smith, A.J.; Lau, J.H.

    1991-01-01

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  20. A compact DD neutron generator-based NAA system to quantify manganese (Mn) in bone in vivo.

    Science.gov (United States)

    Liu, Yingzi; Byrne, Patrick; Wang, Haoyu; Koltick, David; Zheng, Wei; Nie, Linda H

    2014-09-01

    A deuterium-deuterium (DD) neutron generator-based neutron activation analysis (NAA) system has been developed to quantify metals, including manganese (Mn), in bone in vivo. A DD neutron generator with a flux of up to 3*10(9) neutrons s(-1) was set up in our lab for this purpose. Optimized settings, including moderator, reflector, and shielding material and thickness, were selected based on Monte Carlo (MC) simulations conducted in our previous work. Hand phantoms doped with different Mn concentrations were irradiated using the optimized DD neutron generator irradiation system. The Mn characteristic γ-rays were collected by an HPGe detector system with 100% relative efficiency. The calibration line of the Mn/calcium (Ca) count ratio versus bone Mn concentration was obtained (R(2) = 0.99) using the hand phantoms. The detection limit (DL) was calculated to be about 1.05 μg g(-1) dry bone (ppm) with an equivalent dose of 85.4 mSv to the hand. The DL can be reduced to 0.74 ppm by using two 100% HPGe detectors. The whole body effective dose delivered to the irradiated subject was calculated to be about 17 μSv. Given the average normal bone Mn concentration of 1 ppm in the general population, this system is promising for in vivo bone Mn quantification in humans.

  1. The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

    2004-01-01

    The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

  2. Research of the system response of neutron double scatter imaging for MLEM reconstruction

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, M., E-mail: wyj2013@163.com [Northwest Institute of Nuclear Technology, Xi’an 710024 (China); State Key Laboratory of Intense Pulsed Radiation-Simulation and Effect, Xi’an 710024 (China); Peng, B.D.; Sheng, L.; Li, K.N.; Zhang, X.P.; Li, Y.; Li, B.K.; Yuan, Y.; Wang, P.W.; Zhang, X.D.; Li, C.H. [Northwest Institute of Nuclear Technology, Xi’an 710024 (China); State Key Laboratory of Intense Pulsed Radiation-Simulation and Effect, Xi’an 710024 (China)

    2015-03-01

    A Maximum Likelihood image reconstruction technique has been applied to neutron scatter imaging. The response function of the imaging system can be obtained by Monte Carlo simulation, which is very time-consuming if the number of image pixels and particles is large. In this work, to improve time efficiency, an analytical approach based on the probability of neutron interaction and transport in the detector is developed to calculate the system response function. The response function was applied to calculate the relative efficiency of the neutron scatter imaging system as a function of the incident neutron energy. The calculated results agreed with simulations by the MCNP5 software. Then the maximum likelihood expectation maximization (MLEM) reconstruction method with the system response function was used to reconstruct data simulated by Monte Carlo method. The results showed that there was good consistency between the reconstruction position and true position. Compared with back-projection reconstruction, the improvement in image quality was obvious, and the locations could be discerned easily for multiple radiation point sources.

  3. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  4. Study of particle size distribution and formation mechanism of radioactive aerosols generated in high-energy neutron fields

    CERN Document Server

    Endo, A; Noguchi, H; Tanaka, S; Iida, T; Furuichi, S; Kanda, Y; Oki, Y

    2003-01-01

    The size distributions of sup 3 sup 8 Cl, sup 3 sup 9 Cl, sup 8 sup 2 Br and sup 8 sup 4 Br aerosols generated by irradiations of argon and krypton gases containing di-octyl phthalate (DOP) aerosols with 45 MeV and 65 MeV quasi-monoenergetic neutrons were measured in order to study the formation mechanism of radioactive particles in high energy radiation fields. The effects of the size distribution of the radioactive aerosols on the size of the added DOP aerosols, the energy of the neutrons and the kinds of nuclides were studied. The observed size distributions of the radioactive particles were explained by attachment of the radioactive atoms generated by the neutron-induced reactions to the DOP aerosols. (author)

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

    International Nuclear Information System (INIS)

    Janee, H.S.

    1973-09-01

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

  6. Dose equivalent response of personal neutron dosemeters as a function of angle

    International Nuclear Information System (INIS)

    Tanner, J.E.; McDonald, J.C.; Stewart, R.D.; Wernli, C.

    1997-01-01

    The measured and calculated dose equivalent response as a function of angle has been examined for an albedo-type thermoluminescence dosemeter (TLD) that was exposed to unmoderated and D 2 O-moderated 252 Cf neutron sources while mounted on a 40 x 40 15 cm 3 polymethylmethacrylate phantom. The dosemeter used in this study is similar to many neutron personal dosemeters currently in use. The detailed construction of the dosemeter was modelled, and the dose equivalent response was calculated, using the MCNP code. Good agreement was found between the measured and calculated values of the relative dose equivalent angular response for the TLD albedo dosemeter. The relative dose equivalent angular response was also compared with the values of directional and personal dose equivalent as a function of angle published by Siebert and Schuhmacher. (author)

  7. Measurement of neutron spectra generated from bombardment of 4 to 24 MeV protons on a thick 9Be target and estimation of neutron yields

    International Nuclear Information System (INIS)

    Paul, Sabyasachi; Sahoo, G. S.; Tripathy, S. P.; Sunil, C.; Bandyopadhyay, T.; Sharma, S. C.; Ramjilal,; Ninawe, N. G.; Gupta, A. K.

    2014-01-01

    A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK-n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed

  8. Rapid response and wide range neutronic power measuring systems for fast pulsed reactors

    International Nuclear Information System (INIS)

    Sumita, Kenji; Iida, Toshiyuki; Wakayama, Naoaki.

    1976-01-01

    This paper summarizes our investigation on design principles of the rapid, stable and wide range neutronic power measuring system for fast pulsed reactors. The picoammeter, the logarithmic amplifier, the reactivity meter and the neutron current chamber are the items of investigation. In order to get a rapid response, the method of compensation for the stray capacitance of the feedback circuits and the capacitance of signal cables is applied to the picoammeter, the logarithmic amplifier and the reactivity meter with consideration for the stability margin of a whole detecting system. The response of an ionization current chamber and the method for compensating the ion component of the chamber output to get optimum responses high pass filters are investigated. Statistical fluctuations of the current chamber output are also considered in those works. The optimum thickness of the surrounding moderator of the neutron detector is also discussed from the viewpoint of the pulse shape deformation and the neutron sensitivity increase. The experimental results are reported, which were observed in the pulse operations of the one shot fast pulsed reactor ''YAYOI'' and the one shot TRIGA ''NSRR'' with the measuring systems using those principles. (auth.)

  9. Embedded generation for industrial demand response in renewable energy markets

    International Nuclear Information System (INIS)

    Leanez, Frank J.; Drayton, Glenn

    2010-01-01

    Uncertainty in the electrical energy market is expected to increase with growth in the percentage of generation using renewable resources. Demand response can play a key role in giving stability to system operation. This paper discusses the embedded generation for industrial demand response in renewable energy markets. The methodology of the demand response is explained. It consists of long-term optimization and stochastic optimization. Wind energy, among all the renewable resources, is becoming increasingly popular. Volatility in the wind energy sector is high and this is explained using examples. Uncertainty in the wind market is shown using stochastic optimization. Alternative techniques for generation of wind energy were seen to be needed. Embedded generation techniques include co-generation (CHP) and pump storage among others. These techniques are analyzed and the results are presented. From these results, it is seen that investment in renewables is immediately required and that innovative generation technologies are also required over the long-term.

  10. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-11-01

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  11. Neutronic analysis of the Three Mile Island Unit 2 ex-core detector response

    International Nuclear Information System (INIS)

    Malloy, D.J.; Chang, Y.I.

    1981-10-01

    A neutronic analysis has been made with respect to the ex-core neutron detector response during the TMI-2 incident. A series of transport theory calculations quantified the impact upon the detector count rate of various core and downcomer conditions. In particular, various combinations of coolant void content and spatial distributions were investigated to yield the resulting transmission of the photoneutron source to the detector. The impact of a hypothetical distributed source within the downcomer region was also examined in order to simulate the potential effect of the release of neutron producing fission products into the coolant. These results are then offered as potential explanations for the anomalous behavior of the detector during the period of approx. 20 minutes through approx. 3 hours following the reactor scram

  12. Response of CR-39 based personnel neutron dosemeter in terms of directional dose equivalent, in free air and on phantom

    International Nuclear Information System (INIS)

    Pal, Rupali R.; Sathian, Deepa; Jayalakshmi, V.; Chougaonkar, M.P.

    2011-01-01

    CR-39 is the most sensitive of nuclear track detectors for protons and is recommended as an effective neutron dosimeter because of it's low threshold energy of 100 keV neutrons. The fraction of protons that gives detectable tracks in CR-39 depends on the energy of the proton angle of incidence and etching conditions. As a consequence the registration efficiency of neutrons in the CR-39 plastics used for neutron personnel monitoring is strongly influenced by the direction of radiation incidence. This paper presents the relative response of CR-39 at varying neutron incident angles, for 241 Am-Be neutron source spectra in free air and on ISO phantom, in terms of operational quantities. It is observed that the angular dependence of CR-39 for irradiations in air and on phantom is essentially the same indicating that the phantom does not affect the directional response of CR-39. (author)

  13. Study of the prompt gamma ray signal from fissions in special nuclear materials induced using an associated particle neutron generator

    International Nuclear Information System (INIS)

    Koltick, D. S.; Kane, S. Z.

    2009-01-01

    More than 42 million cargo containers entered the United States in 2005. To search for a few kilograms of special nuclear material (SNM) within this vast stream of cargo, an inspection system based on neutron-induced fission followed by the coincident detection of multiple prompt fission gamma rays is investigated using MCNP-Polimi code. The system utilizes two deuterium-tritium (DT) associated particle neutron generators, each capable of 10 9 neutrons/s at 14.1 MeV, with sub-nanosecond timing resolution ZnO:Ga alpha detectors internal to the generator. Because prompt fission signals are approximately 100 times stronger than the delayed signals, the neutron flux is greatly reduced compared to 10 11-12 neutrons/s required for systems based on delayed signals such as the 'nuclear car wash' [4]. In addition the system utilizes 30 cm deep liquid krypton (LKr) noble gas detectors having 94% detection efficiency for 1 MeV gamma rays, high solid angle coverage (∼ 50% of the total solid angle), and sub-nanosecond timing resolution (∼ 600 ps). An algorithm for distinguishing U-235 from U-238 is presented. (authors)

  14. Applicability of the two-angle differential method to response measurement of neutron-sensitive devices at the RCNP high-energy neutron facility

    Energy Technology Data Exchange (ETDEWEB)

    Masuda, Akihiko, E-mail: aki-masuda@aist.go.jp [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Matsumoto, Tetsuro [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Iwamoto, Yosuke [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Hagiwara, Masayuki [High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba, Ibaraki 305-0801 (Japan); Satoh, Daiki; Sato, Tatsuhiko [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Iwase, Hiroshi [High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba, Ibaraki 305-0801 (Japan); Yashima, Hiroshi [Research Reactor Institute, Kyoto University, 2-1010 Asashiro-nishi, Kumatori, Sennan, Osaka 590-0494 (Japan); Nakane, Yoshihiro [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai, Naka, Ibaraki 319-1195 (Japan); Nishiyama, Jun [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo 152-8550 (Japan); Shima, Tatsushi; Tamii, Atsushi; Hatanaka, Kichiji [Research Center for Nuclear Physics (RCNP), Osaka University, 10-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Harano, Hideki [National Institute of Advanced Industrial Science and Technology (AIST), 1-1-1 Umezono, Tsukuba, Ibaraki 305-8568 (Japan); Nakamura, Takashi [Cyclotron and Radioisotope Center (CYRIC), Tohoku University, 6-3 Aramaki, Aoba, Sendai, Miyagi 980-8578 (Japan)

    2017-03-21

    Quasi-monoenergetic high-energy neutron fields induced by {sup 7}Li(p,n) reactions are used for the response evaluation of neutron-sensitive devices. The quasi-monoenergetic high-energy field consists of high-energy monoenergetic peak neutrons and unwanted continuum neutrons down to the low-energy region. A two-angle differential method has been developed to compensate for the effect of the continuum neutrons in the response measurements. In this study, the two-angle differential method was demonstrated for Bonner sphere detectors, which are typical examples of moderator-based neutron-sensitive detectors, to investigate the method's applicability and its dependence on detector characteristics. Experiments were performed under 96–387 MeV quasi-monoenergetic high-energy neutron fields at the Research Center for Nuclear Physics (RCNP), Osaka University. The measurement results for large high-density polyethylene (HDPE) sphere detectors agreed well with Monte Carlo calculations, which verified the adequacy of the two-angle differential method. By contrast, discrepancies were observed in the results for small HDPE sphere detectors and metal-induced sphere detectors. The former indicated that detectors that are particularly sensitive to low-energy neutrons may be affected by penetrating neutrons owing to the geometrical features of the RCNP facility. The latter discrepancy could be consistently explained by a problem in the evaluated cross-section data for the metals used in the calculation. Through those discussions, the adequacy of the two-angle differential method was experimentally verified, and practical suggestions were made pertaining to this method.

  15. The response of the BTI bubble detectors in mixed gamma-neutron workplace fields

    International Nuclear Information System (INIS)

    Vanhavere, F.; Coeck, M.; Lievens, B.; Reginatto, M.

    2005-01-01

    Full text: Bubble detectors have become a mature technology and are used as neutron dosemeters in a wide range of applications. At the SCK-CEN and Belgonucleaire they are used as official personal neutron dosemeter for the personnel. Two types are commercially available from Bubble Technology Industries: the BD-PND, which has a neutron energy threshold of around 100 keV, and the BDT, which is mainly sensitive to thermal neutrons. At Belgonucleaire only the BD-PND is worn, and the results are corrected with a site specific factor. At the SCK-CEN both the BD-PND and BDT are worn and a combination of both results is applied for the dose records. In the EC project EVIDOS (Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields), a whole range of neutron dosemeters were irradiated in workplace fields in nuclear installations in Europe, including both types of bubble detectors. The bubble detectors were exposed on a phantom with different angles towards the reference directions in the workplace fields. We will report the bubble detectors' results in the simulated workplace fields at Cadarache (CANEL and Sigma), in the workplaces at Kruemmel (boiling water reactor, transport cask), at Mol (Venus research reactor SCK-CEN, MOX-fuel facility Belgonucleaire) and Ringhals (pressurized water reactor, transport cask). The responses of the bubble detectors and the combination of both will be compared to the reference values determined with Bonner Spheres and a novel directional spectrometer. The dosemeter readings were checked for consistency by folding the dosemeter response functions with the corresponding workplace fluence spectra in the same workplace. (author)

  16. Relative light yield and temporal response of a stilbene-doped bibenzyl organic scintillator for neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. A.; Goldblum, B. L., E-mail: bethany@nuc.berkeley.edu; Brickner, N. M.; Daub, B. H.; Kaufman, G. S.; Bibber, K. van; Vujic, J. [Department of Nuclear Engineering, University of California, Berkeley, California 94720 (United States); Bernstein, L. A.; Bleuel, D. L.; Caggiano, J. A.; Hatarik, R.; Phillips, T. W.; Zaitseva, N. P. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Wender, S. A. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2014-05-21

    The neutron time-of-flight (nTOF) diagnostics used to characterize implosions at the National Ignition Facility (NIF) has necessitated the development of novel scintillators that exhibit a rapid temporal response and high light yield. One such material, a bibenzyl-stilbene mixed single-crystal organic scintillator grown in a 99.5:0.5 ratio in solution, has become the standard scintillator used for nTOF diagnostics at NIF. The prompt fluorescence lifetime and relative light yield as a function of proton energy were determined to calibrate this material as a neutron detector. The temporal evolution of the intensity of the prompt fluorescent response was modeled using first-order reaction kinetics and the prompt fluorescence decay constant was determined to be 2.46 ± 0.01 (fit) ± 0.13 (systematic) ns. The relative response of the bibenzyl-stilbene mixed crystal generated by recoiling protons was measured, and results were analyzed using Birks' relation to quantify the non-radiative quenching of excitation energy in the scintillator.

  17. Study by Monte Carlo methods of an explosives detection system made up with a D-D neutron generator and NaI(Tl) gamma detectors.

    Science.gov (United States)

    Cevallos Robalino, Lenin E; García Fernández, Gonzalo Felipe; Gallego, Eduardo; Guzmán-García, Karen A; Vega-Carrillo, Hector Rene

    2018-02-17

    Detection of hidden explosives is of utmost importance for homeland security. Several configurations of an Explosives Detection System (EDS) to intercept hidden threats, made up with a Deuterium-Deuterium (D-D) compact neutron generator and NaI (Tl) scintillation detectors, have been evaluated using MCNP6 code. The system's response to various samples of explosives, such as RDX and Ammonium Nitrate, is analysed. The D-D generator is able to produce fast neutrons with 2.5 MeV energy in a maximum yield of 10 10 n/s. It is surrounded by high-density polyethylene to thermalize the fast neutrons and to optimize interactions with the sample inspected, whose emission of gamma rays gives a characteristic spectrum of the elements that constitute it. This procedure allows to determine its chemical composition and to identify the type of substance. The necessary shielding is evaluated to estimate its thicknesses depending on the admissible dose of operation, using lead and polyethylene. The results show that its functionality is promising in the field of national security for explosives inspection. Copyright © 2018 Elsevier Ltd. All rights reserved.

  18. Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Manish K.; Alajo, Ayodeji B., E-mail: alajoa@mst.edu

    2016-08-11

    Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

  19. Design specification for the European Spallation Source neutron generating target element

    International Nuclear Information System (INIS)

    Aguilar, A.; Sordo, F.; Mora, T.; Mena, L.; Mancisidor, M.; Aguilar, J.; Bakedano, G.; Herranz, I.; Luna, P.; Magan, M.; Vivanco, R.; Jimenez-Villacorta, F.; Sjogreen, K.; Oden, U.; Perlado, J.M.

    2017-01-01

    The paper addresses some of the most relevant issues concerning the thermal hydraulics and radiation damage of the neutron generation target to be built at the European Spallation Source as recently approved after a critical design review. The target unit consists of a set of Tungsten blocks placed inside a wheel of 2.5 m diameter which rotates at some 0.5 Hz in order to distribute the heat generated from incoming protons which reach the target in the radial direction. The spallation material elements are composed of an array of Tungsten pieces which rest on a rotating steel support (the cassette) and are distributed in a cross-flow configuration. The thermal, mechanical and radiation effects resulting from the impact of a 2 GeV proton pulse are analysed in detail as well as an evaluation of the inventory of spallation products. The current design is found to conform to specifications and found to be robust enough to deal with several accident scenarios.

  20. Design specification for the European Spallation Source neutron generating target element

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar, A. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Sordo, F., E-mail: fernando.sordo@essbilbao.org [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Instituto de Fusión Nuclear, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Mora, T. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Mena, L. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Instituto de Fusión Nuclear, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Mancisidor, M.; Aguilar, J.; Bakedano, G.; Herranz, I.; Luna, P. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Magan, M.; Vivanco, R. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Instituto de Fusión Nuclear, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Jimenez-Villacorta, F. [Consorcio ESS-BILBAO. Parque Tecnológico Bizkaia. Poligono Ugaldeguren III, Pol. A, 7B, 48170 Zamudio (Spain); Sjogreen, K.; Oden, U. [European Spallation Source ERIC, P.O Box 176, SE-221 00 Lund (Sweden); Perlado, J.M. [Instituto de Fusión Nuclear, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); and others

    2017-06-01

    The paper addresses some of the most relevant issues concerning the thermal hydraulics and radiation damage of the neutron generation target to be built at the European Spallation Source as recently approved after a critical design review. The target unit consists of a set of Tungsten blocks placed inside a wheel of 2.5 m diameter which rotates at some 0.5 Hz in order to distribute the heat generated from incoming protons which reach the target in the radial direction. The spallation material elements are composed of an array of Tungsten pieces which rest on a rotating steel support (the cassette) and are distributed in a cross-flow configuration. The thermal, mechanical and radiation effects resulting from the impact of a 2 GeV proton pulse are analysed in detail as well as an evaluation of the inventory of spallation products. The current design is found to conform to specifications and found to be robust enough to deal with several accident scenarios.