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Sample records for thermal evolution neutron

  1. Magneto–Thermal Evolution of Neutron Stars with Emphasis to ...

    Indian Academy of Sciences (India)

    The magnetic and thermal evolution of neutron stars is a very complex process with many non-linear interactions. For a decent understanding of neutron star physics, these evolutions cannot be considered isolated. A brief overview is presented, which describes the main magneto–thermal interactions that determine the fate ...

  2. Structure and thermal evolution of spinning-down neutron stars

    International Nuclear Information System (INIS)

    Negreiros, R.; Schramm, S.; Weber, F.

    2011-01-01

    In this paper we address the effects of spin-down on the cooling of neutron stars. During its evolution, stellar composition and structure might be substantially altered, as a result of spin-down and the consequent density increase. Since the timescale of cooling might be comparable to to that of the spin-evolution, the modifications to the structure/composition might have important effects on the thermal evolution of the object. We show that the direct Urca process might be delayed or supressed, when spin-down is taken into account. This leads to neutron stars with slow cooling, as opposed to enhanced cooling as would be the case if a "froze-in" structure and composition were considered. In conclusion we demonstrate that the inclusion of spin-down effects on the cooling of neutron stars have far-reaching implications for the interpretation of pulsars. (author)

  3. Statistical approach to thermal evolution of neutron stars

    International Nuclear Information System (INIS)

    Beznogov, M V; Yakovlev, D G

    2015-01-01

    Studying thermal evolution of neutron stars (NSs) is one of a few ways to investigate the properties of superdense matter in their cores. We study the cooling of isolated NSs (INSs) and deep crustal heating of transiently accreting NSs in X-ray transients (XRTs, binary systems with low-mass companions). Currently, nearly 50 of such NSs are observed, and one can apply statistical methods to analyze the whole dataset. We propose a method for such analysis based on thermal evolution theory for individual stars and on averaging the results over NS mass distributions. We calculate the distributions of INSs and accreting NSs (ANSs) in XRTs over cooling and heating diagrams respectively. Comparing theoretical and observational distributions one can infer information on physical properties of superdense matter and on mass distributions of INSs and ANSs. (paper)

  4. Thermal neutron moderating device

    International Nuclear Information System (INIS)

    Takigami, Hiroyuki.

    1995-01-01

    In a thermal neutron moderating device, superconductive coils for generating magnetic fields capable of applying magnetic fields vertical to the longitudinal direction of a thermal neutron passing tube, and superconductive coils for magnetic field gradient for causing magnetic field gradient in the longitudinal direction of the thermal neutron passing tube are disposed being stacked at the outside of the thermal neutron passing tube. When magnetic field gradient is present vertically to the direction of a magnetic moment, thermal neutrons undergo forces in the direction of the magnetic field gradient in proportion to the magnetic moment. Then, the magnetic moment of the thermal neutrons is aligned with the direction vertical to the passing direction of the thermal neutrons, to cause the magnetic field gradient in the passing direction of the thermal neutrons. The speed of the thermal neutrons can be optionally selected and the wavelength can freely be changed by applying forces to the thermal neutrons and changing the extent and direction of the magnetic field gradient. Superconductive coils are used as the coils for generating magnetic fields and the magnetic field gradient in order to change extremely high energy of the thermal neutrons. (N.H.)

  5. Thermal neutron polarisation

    International Nuclear Information System (INIS)

    Satya Murthy, N.S.; Madhava Rao, L.

    1984-01-01

    The basic principle for the production of polarised thermal neutrons is discussed and the choice of various crystal monochromators surveyed. Brief mention of broad-spectrum polarisers is made. The application of polarised neutrons to the study of magnetisation density distributions in magnetic crystals, the dynamic concept of polarisation, principle and use of polarisation analysis, the neutron spin-echo technique are discussed. (author)

  6. Neutron star evolution and emission

    Science.gov (United States)

    Epstein, R. I.; Edwards, B. C.; Haines, T. J.

    1997-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The authors investigated the evolution and radiation characteristics of individual neutron stars and stellar systems. The work concentrated on phenomena where new techniques and observations are dramatically enlarging the understanding of stellar phenomena. Part of this project was a study of x-ray and gamma-ray emission from neutron stars and other compact objects. This effort included calculating the thermal x-ray emission from young neutron stars, deriving the radio and gamma-ray emission from active pulsars and modeling intense gamma-ray bursts in distant galaxies. They also measured periodic optical and infrared fluctuations from rotating neutron stars and search for high-energy TeV gamma rays from discrete celestial sources.

  7. Studsvik thermal neutron facility

    International Nuclear Information System (INIS)

    Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

    1992-01-01

    The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

  8. Semiconductor Thermal Neutron Detector

    Directory of Open Access Journals (Sweden)

    Toru Aoki

    2014-02-01

    Full Text Available The  CdTe  and  GaN  detector  with  a  Gd  converter  have  been developed  and  investigated  as  a  neutron  detector  for neutron  imaging.  The  fabricated  Gd/CdTe  detector  with  the  25  mm  thick  Gd  was  designed  on  the  basis  of  simulation results  of  thermal  neutron  detection  efficiency  and  spatial  resolution.  The  Gd/CdTe  detector  shows  the  detection  of neutron  capture  gamma  ray  emission  in  the  155Gd(n,  g156Gd,  157Gd(n,  g158Gd  and  113Cd(n,  g114Cd  reactions  and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application. Moreover, a BGaN detector has also investigated to separate neutron signal from gamma-ray clearly. 

  9. Advanced plutonium assembly (apa): evolution of the concept, neutron and thermal-mechanic constraints

    International Nuclear Information System (INIS)

    Porta, J.; Gastaldi, B.; Krakowiak-Aillaud, C.; Buffe, L.

    2002-01-01

    The APA concept was developed with the aim of increasing the PWR capacity to burn plutonium emerging from the recycling of irradiated fuels in the French park of nuclear power plants. At first, a concept using annular pins was optimised to allow a good consumption of plutonium while preserving an acceptable neutron control. To cope with the technological problems and those posed by the manufacture of these annular pins, an alternative concept is presented here. It poses as initial conditions the conservation of both the plutonium balance and the respect of the reactivity control. (authors)

  10. Thermal neutron actinide data

    International Nuclear Information System (INIS)

    Tellier, H.

    1992-01-01

    During the 70's, the physicists involved in the cross section measurements for the low energy neutrons were almost exclusively interested in the resonance energy range. The thermal range was considered as sufficiently known. In the beginning of the 80's, reactor physicists had again to deal with the delicate problem of the power reactor temperature coefficient, essentially for the light water reactors. The measured value of the reactivity temperature coefficient does not agree with the computed one. The later is too negative. For obvious safety reasons, it is an important problem which must be solved. Several causes were suggested to explain this discrepancy. Among all these causes, the spectral shift in the thermal energy range seems to be very important. Sensibility calculations shown that this spectral shift is very sensitive to the shape of the neutron cross sections of the actinides for energies below one electron-volt. Consequently, reactor physicists require new and accurate measurements in the thermal and subthermal energy ranges. A part of these new measurement results were recently released and reviewed. The purpose of this study is to complete the preceding review with the new informations which are now available. In reactor physics the major actinides are the fertile nuclei, uranium 238, thorium 232 and plutonium 240 and the fissile nuclei, uranium 233, uranium 235 and plutonium 239. For the fertile nuclei the main datum is the capture cross section, for the fissile nuclei the data of interest are nu-bar, the fission and capture cross sections or a combination of these data such as η or α. In the following sections, we will review the neutron data of the major actinides for the energy below 1 eV

  11. Evolution of Neutron Stars and Observational Constraints

    Directory of Open Access Journals (Sweden)

    Lattimer J.

    2010-10-01

    Full Text Available The structure and evolution of neutron stars is discussed with a view towards constraining the properties of high density matter through observations. The structure of neutron stars is illuminated through the use of several analytical solutions of Einstein’s equations which, together with the maximally compact equation of state, establish extreme limits for neutron stars and approximations for binding energies, moments of inertia and crustal properties as a function of compactness. The role of the nuclear symmetry energy is highlighted and constraints from laboratory experiments such as nuclear masses and heavy ion collisions are presented. Observed neutron star masses and radius limits from several techniques, such as thermal emissions, X-ray bursts, gammaray flares, pulsar spins and glitches, spin-orbit coupling in binary pulsars, and neutron star cooling, are discussed. The lectures conclude with a discusson of proto-neutron stars and their neutrino signatures.

  12. Bibliography for thermal neutron scattering

    International Nuclear Information System (INIS)

    Sakamoto, M.; Chihara, J.; Nakahara, Y.; Kadotani, H.; Sekiya, T.

    1976-12-01

    It contains bibliographical references to measurements, calculations, reviews and basic studies on thermal neutron scatterings and dynamical properties of condensed matter. About 2,700 documents up to the end of 1975 are covered. (auth.)

  13. Evolution of Isolated Neutron Stars

    OpenAIRE

    Popov, S. B.

    2001-01-01

    In this paper we briefly review our recent results on evolution and properties of isolated neutron stars (INSs) in the Galaxy. As the first step we discuss stochastic period evolution of INSs. We briefly discuss how an INS's spin period evolves under influence of interaction with turbulized interstellar medium. To investigate statistical properties of the INS population we calculate a {\\it census} of INSs in our Galaxy. Then we show that for exponential field decay the range of minimum value ...

  14. Thermal neutron source study

    International Nuclear Information System (INIS)

    Holden, T.M.

    1983-05-01

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

  15. In Situ Study of Microstructure Evolution in Solidification of Hypereutectic Al-Si Alloys with Application of Thermal Analysis and Neutron Diffraction

    Science.gov (United States)

    Sediako, Dimitry G.; Kasprzak, Wojciech

    2015-09-01

    Understanding of the kinetics of solid-phase evolution in solidification of hypereutectic aluminum alloys is a key to control their as-cast microstructure and resultant mechanical properties, and in turn, to enhance the service characteristics of actual components. This study was performed to evaluate the solidification kinetics for three P-modified hypereutectic Al-19 pct Si alloys: namely, Al-Si binary alloy and with the subsequent addition of 2.8 pct Cu and 2.8 pct Cu + 0.7 pct Mg. Metallurgical evaluation included thermodynamic calculations of the solidification process using the FactSage™ 6.2 software package, as well as experimental thermal analysis, and in situ neutron diffraction. The study revealed kinetics of solid α-Al, solid Si, Al2Cu, and Mg2Si evolution, as well as the individual effects of Cu and Mg alloying additions on the solidification path of the Al-Si system. Various techniques applied in this study resulted in some discrepancies in the results. For example, the FactSage computations, in general, resulted in 281 K to 286 K (8 °C to 13 °C) higher Al-Si eutectic temperatures than the ones recorded in the thermal analysis, which are also ~278 K (~5 °C) higher than those observed in the in situ neutron diffraction. None of the techniques can provide a definite value for the solidus temperature, as this is affected by the chosen calculation path [283 K to 303 K (10 °C to 30 °C) higher for equilibrium solidification vs non-equilibrium] for the FactSage analysis; and further complicated by evolution of secondary Al-Cu and Mg-Si phases that commenced at the end of solidification. An explanation of the discrepancies observed and complications associated with every technique applied is offered in the paper.

  16. Applications of thermal neutron scattering

    International Nuclear Information System (INIS)

    Kostorz, G.

    1978-01-01

    Although in the past neutrons have been used quite frequently in the study of condensed matter, a more recent development has lead to applications of thermal neutron scattering in the investigation of more practical rather than purely academic problems. Physicists, chemists, materials scientists, biologists, and others have recognized and demonstrated that neutron scattering techniques can yield supplementary information which, in many cases, could not be obtained with other methods. The paper illustrates the use of neutron scattering in these areas of applied research. No attempt is made to present all the aspects of neutron scattering which can be found in textbooks. From the vast amount of experimental data, only a few examples are presented for the study of structure and atomic arrangement, ''extended'' structure, and dynamic phenomena in substances of current interest in applied research. (author)

  17. Measuring thermal neutron characteristics

    International Nuclear Information System (INIS)

    Johnstone, C.W.; Jacobson, L.A.

    1983-01-01

    A method for providing a background-compensated measurement of the level of inducted radiation within an earth formation is claimed. The formation is irradiated with a discrete burst of neutrons and the level of radiation in the formation measured. The level of background radiation is then measured. An average level of both measurements is obtained

  18. Bibliography for thermal neutron scattering

    International Nuclear Information System (INIS)

    Sakamoto, Masanobu; Chihara, Junzo; Gotoh, Yorio; Kadotani, Hiroyuki; Sekiya, Tamotsu.

    1979-09-01

    Bibliographic references are given for measurements, calculations, reviews and basic studies of thermal neutron scattering and dynamical properties of condensed matter. This is the sixth edition covering 3,326 articles collected up to 1978. The edition being the final issue of the present bibliography series, a forthcoming edition will be published in a new form of bibliography. (author)

  19. Fail-safe neutron shutter used for thermal neutron radiography

    International Nuclear Information System (INIS)

    Sachs, R.D.; Morris, R.A.

    1976-11-01

    A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons [neutron flux = 3.876 x 10 6 (neutrons)/(cm 2 .s)]. Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available

  20. Neutron thermalization in light water

    International Nuclear Information System (INIS)

    Abbate, M.J.; Lolich, J.V.

    1975-05-01

    Investigations related to neutron thermalization in light water have been made. Neutron spectra under quasi-infinite-medium conditions have been measured by the time-of-flight technique and calculations were performed with different codes. Through the use of improved experimental techniques and the best known calculational techniques available, the known discrepancies between experimentals and theoretical values were below from 40% to 16%. The present disagreement is believed to be due the scattering model used (ENDF-GASKET, based on the modified Haywood II frequency spectra), that shows to be very satisfactory for poisoned light water cases. Moreover, previous experiments were completed and differential, integral and pulse-source experimental techniques were improved. Also a second step of a neutron and reactor calculation system was completed. (author)

  1. Hyper-thermal neutron irradiation field for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1994-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

  2. Thermal evolution of the CO stretching band in carboxy-myoglobin in the light of neutron scattering and molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Cordone, Lorenzo [Dipartimento di Scienze Fisiche ed Astronomiche, Universita di Palermo and CNISM, Via Archirafi 36, I-90123 Palermo (Italy)], E-mail: cordone@fisica.unipa.it; Cottone, Grazia; Giuffrida, Sergio; Librizzi, Fabio [Dipartimento di Scienze Fisiche ed Astronomiche, Universita di Palermo and CNISM, Via Archirafi 36, I-90123 Palermo (Italy)

    2008-04-18

    As it is well known, the thermal behaviour of the CO stretching band in MbCO reflects the interconversion among protein's taxonomic and lower tier substates. We compare here FTIR data on the thermal behaviour of the CO stretching band in MbCO embedded in non-liquid, water-trehalose matrixes, and neutron scattering data on dry and hydrated proteins and nucleic acids. The comparison, also in the light of simulative data, gives relevant information on the relationship between the mean square displacements of hydrogen atoms and the heme pocket thermal rearrangements in MbCO, as experienced by the bound CO, in the temperature region 100-200 K, and at higher temperature when large scale protein motions take place, following the so-called dynamic transition. The reported results point out how FTIR is a useful tool to study the protein internal dynamics, and complement information from neutron scattering measurements.

  3. Thermal neutron shield and method of manufacture

    Science.gov (United States)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

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

    Science.gov (United States)

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

    2009-06-01

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

  5. Attenuation of thermal neutron through graphite

    International Nuclear Information System (INIS)

    Adib, M.; Ismaail, H.; Fathaallah, M.; Abbas, Y.; Habib, N.; Wahba, M.

    2004-01-01

    Calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of graphite temperature and crystalline from for neutron energies from 1 me V< E<10 eV were carried out. Computer programs have been developed which allow calculation for the graphite hexagonal closed-pack structure in its polycrystalline form and pyrolytic one. I The calculated total cross-section for polycrystalline graphite were compared with the experimental values. An overall agreement is indicated between the calculated values and experimental ones. Agreement was also obtained for neutron cross-section measured for oriented pyrolytic graphite at room and liquid nitrogen temperatures. A feasibility study for use of graphite in powdered form as a cold neutron filter is details. The calculated attenuation of thermal neutrons through large mosaic pyrolytic graphite show that such crystals can be used effectively as second order filter of thermal neutron beams and that cooling improve their effectiveness

  6. Thermal neutron diffusion parameters in homogeneous mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K.; Krynicka, E. [Institute of Nuclear Physics, Cracow (Poland)

    1995-12-31

    A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs.

  7. Thermal neutron diffusion parameters in homogeneous mixtures

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Krynicka, E.

    1995-01-01

    A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs

  8. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  9. Stochastic spin evolution of neutron stars

    OpenAIRE

    Popov, S. B.; Prokhorov, M. E.; Khoperskov, A. V.; Lipunov, V. M.

    2001-01-01

    In this paper we present calculations of period distribution for old accreting isolated neutron stars (INSs). At the age about a few billions years low velocity INSs come to the stage of accretion. At that stage their period evolution is governed by magnetic breaking and accreted angular momentum. Due to turbulence of the interstellar medium (ISM) accreted momentum can both accelerate and decelerate rotation of an INS and spin evolution has chaotic character. Calculations show that for consta...

  10. Attenuation of Thermal Neutrons by Crystalline Silicon

    International Nuclear Information System (INIS)

    Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

    2002-01-01

    A simple formula is given which allows to calculate the contribution of the total neutron cross - section including the Bragg scattering from different (hkt) planes to the neutron * transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy .A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500μ eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

  11. Thermal neutrons streaming in straight duct

    International Nuclear Information System (INIS)

    Jehouani, A.; Boulkheir, M.; Ichaoui, R.

    2000-01-01

    The neutron streaming in duct is due to two phenomena: a) direct propagation and b) reflection on duct wall. We have used the Monte Carlo method to evaluate the ratio of the reflected neutrons flux by the duct wall to the total flux at the exit of the duct for iron and aluminium. Ten neutrons energy groups are considered between 10 -5 eV and 10 eV. A Fortran program is developed to evaluate the neutron double differential albedo. It is shown that the two following approximations are largely justified: i) Three collisions in the duct wall are sufficient to attain the asymptotic limit of the multiscattered neutron double differential albedo ii) The points of entry and exit of the neutron in the duct wall may be considered the same for the multiscattered neutrons. For a punctual source at the mouth of the duct, we have determined the direct and the reflected part of the total thermal neutron flux at the exit of the duct for different lengths and different radius of the duct. For a punctual source, we have found that the major contribution to the total flux of neutrons at the exit is due to the neutron reflection by walls and the reflection contribution decreases when the neutron energy decreases. For a constant length of the duct, the reflected part decreases when the duct radius increases while for the disk shaped source we have found the opposite phenomena. The transmitted neutron flux distribution at the exit of the duct are determined for disk shaped source for different neutron energy and for different distance from the exit center. (author)

  12. Evolution of shell gaps with neutron richness

    International Nuclear Information System (INIS)

    Basu, Moumita Ray; Ray, I.; Kshetri, Ritesh; Saha Sarkar, M.; Sarkar, S.

    2006-01-01

    In the present work, an attempt has been made to coordinate the recent data available over the periodic table, specially near the shell gaps and studied the evolution of the shell gaps as function of neutron numbers and/or other related quantities

  13. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

    Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

  14. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

  15. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

  16. Neutronics methods for thermal radiative transfer

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1988-01-01

    The equations of thermal radiative transfer are time discretized in a semi-implicit manner, yielding a linear transport problem for each time step. The governing equation in this problem has the form of a neutron transport equation with fission but no scattering. Numerical methods are described, whose origins lie in neutron transport, and that have been successfully adapted to this new problem. Acceleration methods that have been developed specifically for the radiative transfer problem, but may have generalizations applicable in neutronics problems, are also discussed

  17. Measurement of thermal neutron capture cross section

    International Nuclear Information System (INIS)

    Huang Xiaolong; Han Xiaogang; Yu Weixiang; Lu Hanlin; Zhao Wenrong

    2001-01-01

    The thermal neutron capture cross sections of 71 Ga(n, γ) 72 Ga, 94 Zr(n, γ) 95 Zr and 191 Ir(n, γ) 192 Ir m1+g,m2 reactions were measured by using activation method and compared with other measured data. Meanwhile the half-life of 72 Ga was also measured. The samples were irradiated with the neutron in the thermal column of heavy water reactor of China Institute of Atomic Energy. The activities of the reaction products were measured by well-calibrated Ge(Li) detector

  18. Thermal neutron albedo measurements for multilithic reflectors

    International Nuclear Information System (INIS)

    Mehboob, Khurram; Ahmed, Raheel; Ali, Majid; Tabassam, Uzma

    2013-01-01

    Highlights: • Measurement of thermal neuron albedo for multilithic reflectors. • Modeling of experiments in MATLAB. • Comparison of numerical calculated and experimental values. • Study of thermal neutron albedo in different multilayered shielding. - Abstract: An experimental measurement of the thermal neutron (0.025 eV) albedo (αth) has been carried out for multilithic shielding by using Am–Be neutron source and BF 3 detector. The measured saturation value for the thermal albedo of paraffin wax has been found to be 0.734 ± 0.020, which is in close agreement to the corresponding value 0.83 quoted in the literature. The thermal neutron albedo has been measured for the multilayered shielding in copper–wood, copper–aluminum, wood–paraffin and paraffin–iron combinations in horizontal geometric configurations. Modeling and numerical simulation have been carried out by developing a MATLAB code which solves the diffusion equation in order to calculate the experimental results. Good agreement has been found between the numerical calculated and experimental results. The uncertainties in the measurements have also been calculated based on error propagation of the underlying Poisson distribution

  19. Method for determining thermal neutron decay times of earth formations

    International Nuclear Information System (INIS)

    Arnold, D.M.

    1976-01-01

    A method is disclosed for measuring the thermal neutron decay time of earth formations in the vicinity of a well borehole. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different intensity modulation frequencies. The tangents of the relative phase angles of the fast neutrons and the resulting thermal neutrons at each of the three frequencies of modulation are measured. First and second approximations to the earth formation thermal neutron decay time are derived from the three tangent measurements. These approximations are then combined to derive a value for the true earth formation thermal neutron decay time

  20. High precision thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Radeka, V.; Schaknowski, N.A.; Smith, G.C.; Yu, B. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    Two-dimensional position sensitive detectors are indispensable in neutron diffraction experiments for determination of molecular and crystal structures in biology, solid-state physics and polymer chemistry. Some performance characteristics of these detectors are elementary and obvious, such as the position resolution, number of resolution elements, neutron detection efficiency, counting rate and sensitivity to gamma-ray background. High performance detectors are distinguished by more subtle characteristics such as the stability of the response (efficiency) versus position, stability of the recorded neutron positions, dynamic range, blooming or halo effects. While relatively few of them are needed around the world, these high performance devices are sophisticated and fairly complex, their development requires very specialized efforts. In this context, we describe here a program of detector development, based on {sup 3}He filled proportional chambers, which has been underway for some years at the Brookhaven National Laboratory. Fundamental approaches and practical considerations are outlined that have resulted in a series of high performance detectors with the best known position resolution, position stability, uniformity of response and reliability over time, for devices of this type.

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

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

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

  2. Thermal and magnetic properties of neutron matter

    International Nuclear Information System (INIS)

    Abd-Alla, M.; Ragab, H.S.; Hassan, M.Y.M.

    1990-01-01

    The Thomas-Fermi model is used to calculate the equation of state of thermal polarized neutron matter applying Seyler-Blanchard interaction. The resulting equation of state is stiff and has a small dependence on both the temperature and the spin excess parameter. We expand the Fermi integrals in powers of temperature up to second order to examine the T 2 approximation for neutron matter. It is found to be reliable up to T = 10 MeV. We also studied the ferromagnetic transition in neutron matter. We found a ferromagnetic transition at density ρ ≅ 2ρ0. This ferromagnetic transition is found to have a small dependence on both the temperature and the spin excess parameter. We also studied the dependence of the effective mass and the sound velocity for polarized neutron matter on temperature. (author). 36 refs, 17 figs

  3. Thermal neutron spectrum distribution in TRIGA fuels

    International Nuclear Information System (INIS)

    Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

    1989-01-01

    The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

  4. Instrumentation to handle thermal polarized neutron beams

    NARCIS (Netherlands)

    Kraan, W.H.

    2004-01-01

    In this thesis we investigate devices needed to handle the polarization of thermal neutron beams: Ï/2-flippers (to start/stop Larmor precession) and Ï-flippers (to reverse polarization/precession direction) and illustrate how these devices are used to investigate the properties of matter and of the

  5. Neutron stars: Observational diversity and evolution

    Science.gov (United States)

    Safi-Harb, S.

    2017-12-01

    Ever since the discovery of the Crab and Vela pulsars in their respective Supernova Remnants, our understanding of how neutron stars manifest themselves observationally has been dramatically shaped by the surge of discoveries and dedicated studies across the electromagnetic spectrum, particularly in the high-energy band. The growing diversity of neutron stars includes the highly magnetized neutron stars (magnetars) and the Central Compact Objects shining in X-rays and mostly lacking pulsar wind nebulae. These two subclasses of high-energy objects, however, seem to be characterized by anomalously high or anomalously low surface magnetic fields (thus dubbed as ‘magnetars’ and ‘anti-magnetars’, respectively), and have pulsar characteristic ages that are often much offset from their associated SNRs’ ages. In addition, some neutron stars act ‘schizophrenic’ in that they occasionally display properties that seem common to more than one of the defined subclasses. I review the growing diversity of neutron stars from an observational perspective, then highlight recent and on-going theoretical and observational work attempting to address this diversity, particularly in light of their magnetic field evolution, energy loss mechanisms, and supernova progenitors’ studies.

  6. The Thermal Neutron Beam Option for NECTAR at MLZ

    Science.gov (United States)

    Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

    The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

  7. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

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

    1979-01-01

    A simultaneous pulsed neutron porosity and thermal neutron capture cross section logging system is provided for radiological well logging of subsurface earth formations. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector, and a combination gamma ray and fast neutron detector is moved through a borehole. Repetitive bursts of neutrons irradiate the earth formations; and, during the bursts, the fast neutron and epithermal neutron populations are sampled. During the interval between bursts the thermal neutron capture gamma ray population is sampled in two or more time intervals. The fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity phi. The capture gamma ray measurements are combined to provide a simultaneous determination of the thermal neutron capture cross section Σ

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

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1995-01-01

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

  9. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

    In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

  10. Molecular evolution and thermal adaptation

    Science.gov (United States)

    Chen, Peiqiu

    2011-12-01

    In this thesis, we address problems in molecular evolution, thermal adaptation, and the kinetics of adaptation of bacteria and viruses to elevated environmental temperatures. We use a nearly neutral fitness model where the replication speed of an organism is proportional to the copy number of folded proteins. Our model reproduces the distribution of stabilities of natural proteins in excellent agreement with experiment. We find that species with high mutation rates tend to have less stable proteins compared to species with low mutation rate. We found that a broad distribution of protein stabilities observed in the model and in experiment is the key determinant of thermal response for viruses and bacteria. Our results explain most of the earlier experimental observations: striking asymmetry of thermal response curves, the absence of evolutionary trade-off which was expected but not found in experiments, correlation between denaturation temperature for several protein families and the Optimal Growth Temperature (OGT) of their carrier organisms, and proximity of bacterial or viral OGTs to their evolutionary temperatures. Our theory quantitatively and with high accuracy described thermal response curves for 35 bacterial species. The model also addresses the key to adaptation is in weak-link genes (WLG), which encode least thermodynamically stable essential proteins in the proteome. We observe, as in experiment, a two-stage adaptation process. The first stage is a Luria-Delbruck type of selection, whereby rare WLG alleles, whose proteins are more stable than WLG proteins of the majority of the population (either due to standing genetic variation or due to an early acquired mutation), rapidly rise to fixation. The second stage constitutes subsequent slow accumulation of mutations in an adapted population. As adaptation progresses, selection regime changes from positive to neutral: Selection coefficient of beneficial mutations scales as a negative power of number of

  11. Cell death following thermal neutron exposure

    Energy Technology Data Exchange (ETDEWEB)

    Paterson, L.C. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Atanackovic, J. [Ontario Power Generation, Toronto, Ontario (Canada); Boyer, C. [Canadian Neutron Beam Centre, Chalk River, Ontario (Canada); El-Jaby, S.; Priest, N.D. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Seymour, C.B.; Boreham, D.R. [McMaster Univ., Hamilton, Ontario (Canada); Richardson, R.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2014-07-01

    When individuals are exposed to unknown external ionizing radiation, it is desirable to have the means to assess both the absorbed dose received (Gy) and the radiation quality. Yet, conventional biodosimetry techniques, specifically the dicentric chromosome assay, cannot differentiate between the damage caused by high- and low-linear energy transfer (LET) exposures. Frequencies of apoptosis and necrosis, may provide an alternative method that assesses both the absorbed dose and radiation quality after unknown exposures. For this preliminary study, human lymphocytes were irradiated with {sup 60}Co gamma rays and thermal neutrons. Both apoptosis and necrosis increased with increasing gamma dose. In contrast, no dose-response was observed following thermal neutron exposure at doses up to 2.61 Gy. (author)

  12. Three frequency modulated combination thermal neutron lifetime log and porosity

    International Nuclear Information System (INIS)

    Paap, H.J.; Arnold, D.M.; Smith, M.P.

    1976-01-01

    Methods are disclosed for measuring simultaneously the thermal neutron lifetime of the borehole fluid and earth formations in the vicinity of a well borehole, together with the formation porosity. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different modulation frequencies. Intensity modulated clouds of thermal neutrons at each of the three modulation frequencies are detected by dual spaced detectors and the relative phase shift of the thermal neutrons with respect to the fast neutrons is determined at each of the three modulation frequencies at each detector. These measurements are then combined to determine simultaneously the thermal neutron decay time of the borehole fluid, the thermal neutron decay time of surrounding earth formation media and the porosity of the formation media

  13. Thermal neutron converter for irradiations with fission neutrons

    International Nuclear Information System (INIS)

    Wagner, F.M.; Kampfer, S.; Kastenmuller, A.; Waschkowski, W.; Bucherl, Th.; Kampfer, S.

    2007-01-01

    The new research reactor FRM II at Garching started operation in March 2004. The compact core is cooled by light water, and moderated by heavy water. Two fuel plates mounted in the heavy water tank convert thermal to fast neutrons. The fast neutron flux in the connected beam tube is up to 7 centre dot 10 8 s -1 cm -2 (depending on filters and collimation); the mean neutron energy is about 1.6 MeV. There are two irradiation rooms along the beam. The first is mainly used for medical therapy (MEDAPP facility), the second for materials characterization (NECTAR facility). At the former therapy facility RENT at the old research reactor FRM, the same beam quality was available until July 2000. Therefore, only a small program is run for the determination of the biological effectiveness of the new beam. The neutron and gamma dose rates in the medical beam are 0.54 and 0.20 Gy/min, respectively. The therapy facility MEDAPP is still under examination according to European regulations for medical devices. Full medical operation will start in 2007. The radiography and tomography facility NECTAR is in operation and aims at non-destructive inspection of objects up to 400 kg mass and 80 centre dot 80 centre dot 80 cm 3 in size. As for fission neutrons the macroscopic cross section of hydrogen is much higher than for other materials (e. g. Fe and Pb), one special application is the detection of hydrogen-containing materials (e. g. oil) in dense materials

  14. Development of temperature related thermal neutron scattering database for MCNP

    International Nuclear Information System (INIS)

    Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

    2013-01-01

    Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

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

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1993-01-01

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

  16. Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

    International Nuclear Information System (INIS)

    Egbert, Stephen D.

    2005-01-01

    In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

  17. maximum neutron flux at thermal nuclear reactors

    International Nuclear Information System (INIS)

    Strugar, P.

    1968-10-01

    Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

  18. Evolution of supernova remnants. III. Thermal waves

    International Nuclear Information System (INIS)

    Chevalier, R.A.

    1975-01-01

    The effect of heat conduction on the evolution of supernova remnants is investigated. A thermal wave, or electron conduction front, can travel more rapidly than a shock wave during the first thousand years of the remnant's evolution. A self-similar solution describing this phase has been found by Barenblatt. Numerical computations verify the solution and give the evolution past the thermal wave phase. While shell formation is not impeded, the interior density and temperature profiles are smoothed by the action of conduction

  19. The measurements of thermal neutron flux distribution in a paraffin

    Indian Academy of Sciences (India)

    The term `thermal flux' implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 ms-1 at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of ...

  20. Analytic scattering kernels for neutron thermalization studies

    International Nuclear Information System (INIS)

    Sears, V.F.

    1990-01-01

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

  1. Maximum neutron flux in thermal reactors

    International Nuclear Information System (INIS)

    Strugar, P.V.

    1968-12-01

    Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

  2. Physics with Ultracold and Thermal Neutron Beams

    International Nuclear Information System (INIS)

    None

    2004-01-01

    The final report is broken into 5 segments, reflecting research conclusions reached during specific time periods: 1991-1997, 1997-1999, 1999-2000, 2000-2001, and 2001-2002. The first part of the work reported was carried out at the 2 Mw research reactor of the Rhode Island Nuclaer Science Center (RJNSC). Chosen for study was the slow phase separation in mixtures of oil and water in the presence of a surfactant, and the structural features of an oil layer during the slow build-up from the gas phase. The results of these measurements, as well as studies of the capillary wave properties of oil/surfactant/water interfaces are described. The second part of the work was performed at the neutron reflection facilities of the Intennse Pulsed Neutron Source at Argonne and of the NBSR reactor at NIST. At Argonne, the uniaxial magnetic order of an Fe/CR superlattice was investigated, while the experiments at NIST studied the swelling behavior of ordered thin films of diblock copolymers when they were exposed to solvent vapors. The third part of the work was concerned with the storage properties of ultracold neturons in a trap. New experiments on spectral evolution during storage, using the UCN source of the Institut Laue-Langevin were able to be run. Subsequent periods focussed on the ultracold neutrons work, spin valve multilayer systems, and pseudo-partial wetting

  3. Thermalization of monoenergetic neutrons in a concrete room

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Iniguez, M.P.; Martin M, A. [Universidad de Valladolid, (Spain)

    2006-07-01

    The thermalization of neutrons from monoenergetic neutron sources in a concrete room has been studied. During calibration of neutron detectors it is mandatory to make corrections due to neutron scattering produced by the room walls, therefore this factor must be known in advance. The scattered neutrons are thermalized and produce a neutron field that is directly proportional to source strength and inversely proportional to room total wall-surfaces, the proportional coefficient has been calculated for neutrons whose energy goes from 1 eV to 20 MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 and 300 cm-radius spherical cavity, where monoenergetic neutrons were located at the center, along the spherical cavity radius neutron spectra were calculated at several source-to-detector distances inside the cavity. The obtained coefficient is almost three times larger than the factor normally utilized. (Author)

  4. Neutrino flavor evolution in neutron star mergers

    Science.gov (United States)

    Tian, James Y.; Patwardhan, Amol V.; Fuller, George M.

    2017-08-01

    We examine the flavor evolution of neutrinos emitted from the disklike remnant (hereafter called "neutrino disk") of a binary neutron star (BNS) merger. We specifically follow the neutrinos emitted from the center of the disk, along the polar axis perpendicular to the equatorial plane. We carried out two-flavor simulations using a variety of different possible initial neutrino luminosities and energy spectra and, for comparison, three-flavor simulations in specific cases. In all simulations, the normal neutrino mass hierarchy was used. The flavor evolution was found to be highly dependent on the initial neutrino luminosities and energy spectra; in particular, we found two broad classes of results depending on the sign of the initial net electron neutrino lepton number (i.e., the number of neutrinos minus the number of antineutrinos). In the antineutrino-dominated case, we found that the matter-neutrino resonance effect dominates, consistent with previous results, whereas in the neutrino-dominated case, a bipolar spectral swap develops. The neutrino-dominated conditions required for this latter result have been realized, e.g., in a BNS merger simulation that employs the "DD2" equation of state for neutron star matter [Phys. Rev. D 93, 044019 (2016), 10.1103/PhysRevD.93.044019]. For this case, in addition to the swap at low energies, a collective Mikheyev-Smirnov-Wolfenstein mechanism generates a high-energy electron neutrino tail. The enhanced population of high-energy electron neutrinos in this scenario could have implications for the prospects of r -process nucleosynthesis in the material ejected outside the plane of the neutrino disk.

  5. Thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Tuli, J.K.

    1983-01-01

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

  6. More accurate thermal neutron coincidence counting technique

    International Nuclear Information System (INIS)

    Baron, N.

    1978-01-01

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

  7. Applications of thermal neutron scattering in biology, biochemistry and biophysics

    International Nuclear Information System (INIS)

    Worcester, D.L.

    1977-01-01

    Biological applications of thermal neutron scattering have increased rapidly in recent years. The following categories of biological research with thermal neutron scattering are presently identified: crystallography of biological molecules; neutron small-angle scattering of biological molecules in solution (these studies have already included numerous measurements of proteins, lippoproteins, viruses, ribosomal subunits and chromatin subunit particles); neutron small-angle diffraction and scattering from biological membranes and membrane components; and neutron quasielastic and inelastic scattering studies of the dynamic properties of biological molecules and materials. (author)

  8. Real-time thermal neutron radiographic detection systems

    International Nuclear Information System (INIS)

    Berger, H.; Bracher, D.A.

    1976-01-01

    Systems for real-time detection of thermal neutron images are reviewed. Characteristics of one system are presented; the data include contrast, resolution and speed of response over the thermal neutron intensity range 2.5 10 3 n/cm 2 -sec to 10 7 n/cm 2 -sec

  9. Evolution of newborn neutron stars: role of quark matter nucleation

    International Nuclear Information System (INIS)

    Bombaci, Ignazio; Logoteta, Domenico; Providencia, Constança; Vidaña, Isaac

    2011-01-01

    A phase of strong interacting matter with deconfined quarks is expected in the core of massive neutron stars. We study the quark deconfinement phase transition in cold (T = 0) and hot β-stable hadronic matter. Assuming a first order phase transition, we calculate and compare the nucleation rate and the nucleation time due to thermal and quantum nucleation mechanisms. We show that above a threshold value of the central pressure a pure hadronic star (HS) is metastable to the conversion to a quark star (QS) (i.e. hybrid star or strange star). We introduce the concept of critical mass M cr for cold HSs and proto-hadronic stars (PHSs), and the concept of limiting conversion temperature for PHSs. We show that PHSs with a mass M cr could survive the early stages of their evolution without decaying to QSs. Finally, we discuss the possible evolutionary paths of proto-hadronic stars.

  10. Neutrino opacities in kaon condensation and evolution of neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Muto, Takumi [Chiba Institute of Technology, Dept. of Physics, Narashino, Chiba (Japan); Yasuhira, Masatomi [Kyoto Univ., Yukawa Institute for Theoretical Physics, Kyoto (Japan); Tatsumi, Toshitaka [Kyoto Univ., Dept. of Physics, Kyoto (Japan); Iwamoto, Naoki [Kagawa Univ., Faculty of Engineering, Takamatsu, Kagawa (Japan)

    2002-09-01

    The neutrino mean free paths are obtained in kaon condensates realized from hot neutron-star matter. Kaon-induced neutrino absorption processes (KA), {nu}{sub e}N {yields} e{sup -}N (N stands for the nucleon), which are unique in the presence of kaon condensates, are mainly considered in nondegenerate neutrino case. The mean free paths for the KA processes are compared with the neutrino scatterings (S), {nu}{sub e}N {yields} {nu}{sub e}N. It is shown that the mean free paths for KA are shorter than the ordinary two-nucleon processes, {nu}{sub e}nN {yields} e{sup -}pN by several orders of magnitude when the temperature is not very high. However, the scattering processes have a dominant contribution to the neutrino opacities as compared with KA, so that KA has a minor effect on the thermal and dynamical evolution of protoneutron stars. (author)

  11. Design of hyper-thermal neutron irradiation fields for neutron capture therapy in KUR-heavy water neutron irradiation facility. Mounting of hyper-thermal neutron converter in therapeutic collimator

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2001-01-01

    Neutron capture therapy (NCP) using thermal neutron needs to improve of depth dose distribution in a living body. Epi-thermal neutron following moderation of fast neutron is usually used for improving of the depth dose distribution. The moderation method of fast neutron, however, gets mixed some of high energy neutron which give some of serious effects to a living body, and involves the difficulty for collimation of thermal neutron to the diseased part. Hyper-thermal neutrons, which are in an energy range of 0.1-3 eV at high temperature side of thermal neutron, are under consideration for application to the NCP. The hyper-thermal neutrons can be produced by up-scattering of thermal neutron in a high temperature material. Fast neutron components in collimator for the NCP reduce on application of the up-scattering method. Graphite at high temperature (>1000k) is used as a hyper-thermal neutron converter. The hyper-thermal neutron converter is planted to mount on therapeutic collimator which is located at the nearest side of patient for the NCP. Total neutron flux, ratio of hyper-thermal neutron to total neutron, and ratio of gamma-ray dose to neutron flux are calculated as a function of thickness of the graphite converter using monte carlo code MCNP-V4B. (M. Suetake)

  12. Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

    International Nuclear Information System (INIS)

    Nemec, T.; Rant, J.; Kristof, E.; Glumac, B.

    1995-01-01

    An extensive characterization of the neutron beam of the existing neutron radiographic facility in the thermal column of the Ljubljana Triga Mark II research reactor is in progress. Neutron beam characteristics are needed to determine the effect of various neutron and gamma radiation on the neutron radiographic image. Commercially available medical scintillator converter screens based on Gd dioxy sulphite as well as Gd metal neutron converters are used to record neutron radiographic image. Thermal, epithermal and fast neutron fluxes were measured using Au and In activation detectors and cadmium ratio is determined. Neutron beam flux profiles are measured by film densitometry and by Au activation detector wires. By exposing films shielded by boral or lead plates individual contributions of thermal, epithermal neutrons and gamma radiation are estimated by densitometric measurements. By recording images of neutron image quality indicators BPI (Beam Purity Indicator) and SI (Sensitivity Indicator) produced by Riso, standard neutron radiography image characteristic are established. In gamma dosimetric measurements thermoluminescent detectors (CaF 2 Mn) are used. (author)

  13. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  14. Hydrodynamic evolution of neutron star merger remnants

    Science.gov (United States)

    Liu, Men-Quan; Zhang, Jie

    2017-11-01

    Based on the special relativistic hydrodynamic equations and updated cooling function, we investigate the long-term evolution of neutron stars merger (NSM) remnants by a one-dimensional hydrodynamic code. Three NSM models from one soft equation of state, SFHo, and two stiff equations of state, DD2 and TM1, are used to compare their influences on the hydrodynamic evolution of remnants. We present the luminosity, mass and radius of remnants, as well as the velocity, temperature and density of shocks. For a typical interstellar medium (ISM) density with solar metallicity, we find that the NSM remnant from the SFHo model makes much more changes to ISM in terms of velocity, density and temperature distributions, compared with the case of DD2 and TM1 models. The maximal luminosity of the NSM remnant from the SFHo model is 3.4 × 1038 erg s-1, which is several times larger than that from DD2 and TM1 models. The NSM remnant from the SFHo model can maintain high luminosity (>1038 erg s-1) for 2.29 × 104 yr. Furthermore, the density and temperature of remnants at the maximal luminosity are not sensitive to the power of the original remnant. For the ISM with the solar metallicity and nH = 1 cm- 3, the density of the first shock ∼10-23 g cm-3 and the temperature ∼3 × 105 K in the maximal luminosity phase; The temperature of the first shock decreases and there is a thin 'dense' shell with density ∼10-21 g cm-3 after the maximal luminosity. These characteristics may be helpful for future observations of NSM remnants.

  15. Using thermalizers in measuring 'Ukryttia' object's FCM neutron fluxes

    CERN Document Server

    Krasnyanskaya, O G; Odinokin, G I; Pavlovich, V N

    2003-01-01

    The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutron flux measuring in the 'Ukryttia' are described. The calculations of neutron flux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutron flux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

  16. Thermal neutron standard fields with the KUR heavy water facility

    International Nuclear Information System (INIS)

    Kanda, K.; Kobayashi, K.; Shibata, T.

    1978-01-01

    A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure thermal neutrons in the Maxwellian distribution. The facility is faced to the core of KUR and it contains about 2 tons of heavy water. The thickness of the layer is about 140 cm. The neutron spectrum was measured with the time of flight technique using a fast chopper. The measured spectrum was in good agreement with the Maxwellian distribution in all energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons of the γ-n reaction of heavy water was very small. The maximum intensity of thermal neutrons is 3x10 11 n/cm 2 sec. When the bismuth scatterer is attached, the gamma rays contamination is eliminated by the ratio of 0.05 of gamma rays to neutrons in rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, detector calibration, activation analysis, biomedical purposes etc. (author)

  17. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

    Shultz, W.E.

    1980-01-01

    A method for simultaneously determining the porosity and thermal neutron capture cross-section of earth formations in the vicinity of a well borehole is claimed. It comprises the following steps: passing a well tool into a cased well borehole. The tool has a pulsed source of fast neutrons, a combination fast neutron and gamma ray detector and an epithermal neutron detector; repetitively irradiating the earth formations in the vicinity of the borehole with bursts of fast neutrons; detecting the fast neutron and epithermal neutron populations in the borehole (during the neutron bursts) and generating first and second measurement signals; detecting for second and third time intervals during the time between the neutron bursts, the gamma radiation present in the borehole due to the capture of thermalized neutrons by the nuclei of elements comprising the earth formations and generating third and fourth measurement signals; and combining the first and second measurement signals according to a predetermined relationship to derive an indication of the porosity of the earth formations and combining the third and fourth measurement signals to derive an indication of the thermal neutron capture cross-section of the earth formations

  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. Thermal evolution of compact stars

    International Nuclear Information System (INIS)

    Schaab, C.; Glendenning, N.K.

    1996-01-01

    A collection of modern, field-theoretical equations of state is applied to the investigation of cooling properties of compact stars. These comprise neutron stars as well as hypothetical strange-matter stars, made up of absolutely stable 3-flavor strange-quark matter. Various uncertainties in the behavior of matter at supernuclear densities, e.g., hyperonic degrees of freedom, behavior of coupling strengths in matter, pion and meson condensation, superfluidity, transition to quark matter, absolute stability of strange-quark matter, and last but not least the many-body technique itself are tested against the body of observed cooling data. (orig.)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  1. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

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

    1980-01-01

    An improved method and apparatus are described for simultaneously measuring the porosity and thermal neutron capture cross section of earth formations in situ in the vicinity of a well borehole using pulsed neutron well logging techniques. The logging tool which is moved through the borehole consists of a 14 MeV pulsed neutron source, an epithermal neutron detector and a combination gamma ray and fast neutron detector. The associated gating systems, counters and combined digital computer are sited above ground. (U.K.)

  2. Neutron capture therapy with thermal neutrons at IRT MIFI

    International Nuclear Information System (INIS)

    Zajtsev, K.N.; Portnov, A.A.; Savkin, V.A.; Kulakov, V.N.; Khokhlov, V.F.; Shejno, I.N.; Vajnson, A.A.; Kozlovskaya, N.G.; Meshcherikova, V.V.; Mitin, V.N.; Yarmonenko, S.P.

    2001-01-01

    Combined preclinical investigations into neutron capture therapy are conducted. Malignant melanoma was adopted as the line of investigation; boron-containing and gadolinium-containing preparations were used during the neutron capture therapy working off. Preparations produce secondary varying radiations when used in tumor. Dogs with spontaneous melanoma were used for the experiments. Procedures for the irradiation of dogs by neutron beam as the stage before use for the treatment of oncology patients were finished off; efficiency of neutron beam influence on normal tissues during the irradiation of dogs with melanoma (and without it) in antitumor and side effect sense was estimated [ru

  3. Characteristics of thermal neutron calibration fields using a graphite pile

    International Nuclear Information System (INIS)

    Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

    2005-03-01

    The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

  4. Stereoscopic radiographic images with thermal neutrons

    International Nuclear Information System (INIS)

    Silvani, M.I.; Almeida, G.L.; Rogers, J.D.; Lopes, R.T.

    2011-01-01

    Spatial structure of an object can be perceived by the stereoscopic vision provided by eyes or by the parallax produced by movement of the object with regard to the observer. For an opaque object, a technique to render it transparent should be used, in order to make visible the spatial distribution of its inner structure, for any of the two approaches used. In this work, a beam of thermal neutrons at the main port of the Argonauta research reactor of the Instituto de Engenharia Nuclear in Rio de Janeiro/Brazil has been used as radiation to render the inspected objects partially transparent. A neutron sensitive Imaging Plate has been employed as a detector and after exposure it has been developed by a reader using a 0.5 μm laser beam, which defines the finest achievable spatial resolution of the acquired digital image. This image, a radiographic attenuation map of the object, does not represent any specific cross-section but a convoluted projection for each specific attitude of the object with regard to the detector. After taking two of these projections at different object attitudes, they are properly processed and the final image is viewed by a red and green eyeglass. For monochromatic images this processing involves transformation of black and white radiographies into red and white and green and white ones, which are afterwards merged to yield a single image. All the processes are carried out with the software ImageJ. Divergence of the neutron beam unfortunately spoils both spatial and contrast resolutions, which become poorer as object-detector distance increases. Therefore, in order to evaluate the range of spatial resolution corresponding to the 3D image being observed, a curve expressing spatial resolution against object-detector gap has been deduced from the Modulation Transfer Functions experimentally. Typical exposure times, under a reactor power of 170 W, were 6 min for both quantitative and qualitative measurements. In spite of its intrinsic constraints

  5. Stereoscopic radiographic images with thermal neutrons

    Science.gov (United States)

    Silvani, M. I.; Almeida, G. L.; Rogers, J. D.; Lopes, R. T.

    2011-10-01

    Spatial structure of an object can be perceived by the stereoscopic vision provided by eyes or by the parallax produced by movement of the object with regard to the observer. For an opaque object, a technique to render it transparent should be used, in order to make visible the spatial distribution of its inner structure, for any of the two approaches used. In this work, a beam of thermal neutrons at the main port of the Argonauta research reactor of the Instituto de Engenharia Nuclear in Rio de Janeiro/Brazil has been used as radiation to render the inspected objects partially transparent. A neutron sensitive Imaging Plate has been employed as a detector and after exposure it has been developed by a reader using a 0.5 μm laser beam, which defines the finest achievable spatial resolution of the acquired digital image. This image, a radiographic attenuation map of the object, does not represent any specific cross-section but a convoluted projection for each specific attitude of the object with regard to the detector. After taking two of these projections at different object attitudes, they are properly processed and the final image is viewed by a red and green eyeglass. For monochromatic images this processing involves transformation of black and white radiographies into red and white and green and white ones, which are afterwards merged to yield a single image. All the processes are carried out with the software ImageJ. Divergence of the neutron beam unfortunately spoils both spatial and contrast resolutions, which become poorer as object-detector distance increases. Therefore, in order to evaluate the range of spatial resolution corresponding to the 3D image being observed, a curve expressing spatial resolution against object-detector gap has been deduced from the Modulation Transfer Functions experimentally. Typical exposure times, under a reactor power of 170 W, were 6 min for both quantitative and qualitative measurements. In spite of its intrinsic constraints

  6. GEM-based thermal neutron beam monitors for spallation sources

    International Nuclear Information System (INIS)

    Croci, G.; Claps, G.; Caniello, R.; Cazzaniga, C.; Grosso, G.; Murtas, F.; Tardocchi, M.; Vassallo, E.; Gorini, G.; Horstmann, C.; Kampmann, R.; Nowak, G.; Stoermer, M.

    2013-01-01

    The development of new large area and high flux thermal neutron detectors for future neutron spallation sources, like the European Spallation Source (ESS) is motivated by the problem of 3 He shortage. In the framework of the development of ESS, GEM (Gas Electron Multiplier) is one of the detector technologies that are being explored as thermal neutron sensors. A first prototype of GEM-based thermal neutron beam monitor (bGEM) has been built during 2012. The bGEM is a triple GEM gaseous detector equipped with an aluminum cathode coated by 1μm thick B 4 C layer used to convert thermal neutrons to charged particles through the 10 B(n, 7 Li)α nuclear reaction. This paper describes the results obtained by testing a bGEM detector at the ISIS spallation source on the VESUVIO beamline. Beam profiles (FWHM x =31 mm and FWHM y =36 mm), bGEM thermal neutron counting efficiency (≈1%), detector stability (3.45%) and the time-of-flight spectrum of the beam were successfully measured. This prototype represents the first step towards the development of thermal neutrons detectors with efficiency larger than 50% as alternatives to 3 He-based gaseous detectors

  7. Thermal neutron detectors based on complex oxide crystals

    CERN Document Server

    Ryzhikov, V; Volkov, V; Chernikov, V; Zelenskaya, O

    2002-01-01

    The ways of improvement of spectrometric quality of CWO and GSO crystals have been investigated with the aim of their application in thermal neutron detectors based on radiation capture reactions. The efficiency of the neutron detection by these crystals was measured, and the obtained data were compared with the results for sup 6 LiI(Tl) crystals. It is shown that the use of complex oxide crystals and neutron-absorption filters for spectrometry of thermal and resonance neutrons could be a promising method in combination with computer data processing. Numerical calculations are reported for spectra of gamma-quanta due to radiation capture of the neutrons. To compensate for the gamma-background lines, we used a crystal pair of heavy complex oxides with different sensitivity to neutrons.

  8. Evolution of Neutron Star Magnetic Fields

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    in nuclei. The neutrons are expected to form a 3P superfluid and the protons a 1S ... crust are expected to form a lattice; the electrons are free and highly degenerate, .... the reduced magnetic fields in neutron stars processed in binaries,.

  9. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    International Nuclear Information System (INIS)

    Vaiana, F.

    2009-11-01

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  10. Neutron thermalization and spectra; Thermalisation et spectres de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Cadilhac, M; Soule, J L; Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The investigation of the neutron spectra in thermal reactors is taking an increasing importance due to the role played in them by Plutonium. Whatever the absorption law, it has been remarked that the scattering law in the.moderator affects the spectrum only through certain overall properties. It would thus seem possible to develop a simplified representation of this effect which would lead to a clear understanding of the phenomena, reducing at the same time the volume of numerical calculations required.. The synthetic model employed by the authors presents the advantage of reducing the determination of the spectra in an homogeneous medium to the resolution of a second order differential equation, like the Wigner-Wilkins model (monoatomic gaseous hydrogen) and the generalized heavy gas model of J. Horowitz which, incidentally, are both special cases. The model is, on the other hand, sufficiently general to allow a correct treatment of the situations met with in practice and in particular the important case where the presence of Plutonium introduces absorption resonances at low energy. Actually, the chemical or crystalline bonds of the moderator are introduced into the proposed model through two energy functions. These functions have been adjusted for the usual moderators (Graphite heavy water, light water) by means of known theoretical scattering laws. In a heterogeneous medium, the most important factor is the mean spectrum in the fuel of one cell, the knowledge of which is allowed by a generalization of the Amouyal-Benoist-Horowitz method. The proposed model lends itself particularly well to such calculations and also allows the effects of re-thermalization (for instance when the cooling system and the moderator are at different temperatures) to be treated. Finally, some examples are given of practical applications: codes for spectra and effective cross sections computations (editing of tables), codes for the treatment of neutron balance in a lattice or for the

  11. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  12. Monte Carlo calculations of neutron thermalization in a heterogeneous system

    Energy Technology Data Exchange (ETDEWEB)

    Hoegberg, T

    1959-07-15

    The slowing down of neutrons in a heterogeneous system (a slab geometry) of uranium and heavy water has been investigated by Monte Carlo methods. Effects on the neutron spectrum due to the thermal motions of the scattering and absorbing atoms are taken into account. It has been assumed that the speed distribution of the moderator atoms are Maxwell-Boltzmann in character.

  13. Determination of average activating thermal neutron flux in bulk samples

    International Nuclear Information System (INIS)

    Doczi, R.; Csikai, J.; Doczi, R.; Csikai, J.; Hassan, F. M.; Ali, M.A.

    2004-01-01

    A previous method used for the determination of the average neutron flux within bulky samples has been applied for the measurements of hydrogen contents of different samples. An analytical function is given for the description of the correlation between the activity of Dy foils and the hydrogen concentrations. Results obtained by the activation and the thermal neutron reflection methods are compared

  14. Calculation of the neutron parameters of fast thermal reactor

    International Nuclear Information System (INIS)

    Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

    1975-01-01

    The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

  15. The determination of thermal neutron cross section of 81Br

    International Nuclear Information System (INIS)

    Kovacs, Luciana; Zamboni, Cibele B.; Dalaqua Junior, Leonardo

    2009-01-01

    In this investigation several standard materials were used to determine the thermal neutron cross section of 81 Br. This nuclear parameter is an important data to perform several quantitative investigations, mainly in medical area. In other to confirm and to reduce the uncertainty, a new measurement was preformed using thermal neutron at IEA-R1 nuclear reactor of IPEN/CNEN-SP. The result obtained is compatible with the tabulated value and present small uncertainly. (author)

  16. Thermal neutron capture cross sections of tellurium isotopes

    International Nuclear Information System (INIS)

    Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

    2003-01-01

    New values for thermal neutron capture cross sections of the tellurium isotopes 122 Te, 124 Te, 125 Te, 126 Te, 128 Te, and 130 Te are reported. These values are based on a combination of newly determined partial γ-ray cross sections obtained from experiments on targets contained natural Te and γ intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

  17. Thermal neutron capture cross sections of tellurium isotopes

    International Nuclear Information System (INIS)

    Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

    2004-01-01

    New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

  18. Thermal neutron capture cross sections of tellurium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Tomandl, I.; Honzatko, J.; von Egidy, T.; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

    2004-03-01

    New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given.

  19. Study and development of new dosemeters for thermal neutrons

    International Nuclear Information System (INIS)

    Urena N, F.

    1998-01-01

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  20. Thermal neutron imaging in an active interrogation environment

    International Nuclear Information System (INIS)

    Vanier, P.E.; Forman, L.; Norman, D.R.

    2009-01-01

    We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of excitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

  1. Manufacturing of thermal neutron sensor using pMOS

    International Nuclear Information System (INIS)

    Lee, Nam Ho; Kim, Seung Ho

    2005-05-01

    A pMOSFET sensor having a Gadolinium converter has been invented successfully as a slow neutron sensor that is sensitive to neutron energy down to 0.025 eV. The Gd layer converts low energy neutrons to ionizing radiation of which the amount is proportional to neutron dose. Ionising radiation from neutron reactions changes the charge state of the gate oxide of the pMOSFET. The Gd-pMOSFETs were tested at a neutron beam port of HANARO research reactor and a 60 CO irradiation facility to investigate slow neutron response and gamma response, respectively. The voltage change was proportional to the accumulated slow neutron dose. The results from Gd coupled MOSFET neutron dosemeters shows an excellent sensitivity (15 - 16mV/cGy) and linearity to thermal neutrons with negligible background contamination. The results demonstrate the outstanding performance of the Gd coupled MOSFET neutron dosemeters clearly. The Gd-pMOSFET can also be used in a mixed radiation field by subtracting the voltage change of a pMOSFET without Gd from that of the Gd-pMOSFET

  2. Study of a nTHGEM-based thermal neutron detector

    Science.gov (United States)

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

    2016-07-01

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

  3. Thermal neutron flux measurements using neutron-electron converters; Mesure de flux de neutrons thermiques avec des convertisseurs neutrons electrons

    Energy Technology Data Exchange (ETDEWEB)

    Le Meur, R; Lecomte, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    The operation of neutron-electron converters designed for measuring thermal neutron fluxes is examined. The principle is to produce short lived isotopes emitting beta particles, by activation, and to measure their activity not by extracting them from the reactor, but directly in the reactor using the emitted electrons to deflect the needle of a galvanometer placed outside the flux. After a theoretical study, the results of the measurements are presented; particular attention is paid to a new type of converter characterized by a layer structure. The converters are very useful for obtaining flux distributions with more than 10{sup 7} neutrons cm{sup -2}*sec{sup -1}. They work satisfactorily in pressurized carbon dioxide at 400 Celsius degrees. Some points still have to be cleared up however concerning interfering currents in the detectors and the behaviour of the dielectrics under irradiation. (authors) [French] On examine le fonctionnement de convertisseurs neutrons electrons destines a des mesures de flux de neutrons thermiques. Le principe est de former par activation des isotopes a periodes courtes et a emission beta et de mesurer leur activite non pas en les sortant du reacteur, mais directement en pile, utilisant les electrons emis pour faire devier l'aiguille d'un galvanometre place hors flux. Apres une etude theorique, on indique des resultats de mesures obtenus, en insistant particulierement sur un nouveau type de convertisseur, caracterise par sa structure stratifiee. Les convertisseurs sont tres interessants pour tracer, des cartes de flux a partir de 10{sup 7} neutrons cm{sup -2}*s{sup -1}. Ils sont utilisables pour des flux de 10{sup 14} neutrons cm{sup -2}*s{sup -1}. Ils fonctionnent correctement dans du gaz carbonique sous pression a 400 C. Des points restent cependant a eclaircir concernant les courants parasites dans les detecteurs et le comportement des dielectriques pendant leur irradiation. (auteur)

  4. Experimental characterization of semiconductor-based thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Bortot, D.; Pola, A.; Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN—Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sacco, D. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); INAIL—DIT, Via di Fontana Candida 1, 00040 Monteporzio Catone (Italy); Esposito, A.; Gentile, A.; Buonomo, B. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Palomba, M.; Grossi, A. [ENEA Triga RC-1C.R. Casaccia, via Anguillarese 301, 00060 S. Maria di Galeria, Roma (Italy)

    2015-04-21

    In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of {sup 6}LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62 μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 10{sup 12} cm{sup −2})

  5. Accounting for the thermal neutron flux depression in voluminous samples for instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Overwater, R.M.W.; Hoogenboom, J.E.

    1994-01-01

    At the Delft University of Technology Interfaculty Reactor Institute, a facility has been installed to irradiate cylindrical samples with diameters up to 15 cm and weights up to 50 kg for instrumental neutron activation analysis (INAA) purposes. To be able to do quantitative INAA on voluminous samples, it is necessary to correct for gamma-ray absorption, gamma-ray scattering, neutron absorption, and neutron scattering in the sample. The neutron absorption and the neutron scattering are discussed. An analytical solution is obtained for the diffusion equation in the geometry of the irradiation facility. For samples with known composition, the neutron flux--as a function of position in the sample--can be calculated directly. Those of unknown composition require additional flux measurements on which least-squares fitting must be done to obtain both the thermal neutron diffusion coefficient D s and the diffusion length L s of the sample. Experiments are performed to test the theory

  6. Neutron moderation theory with thermal motion of the moderator nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Rusov, V.D.; Tarasov, V.A.; Chernezhenko, S.A.; Kakaev, A.A.; Smolyar, V.P. [Odessa National Polytechnic University, Department of Theoretical and Experimental Nuclear Physics, Odessa (Ukraine)

    2017-09-15

    In this paper we present the analytical expression for the neutron scattering law for an isotropic source of neutrons, obtained within the framework of the gas model with the temperature of the moderating medium as a parameter. The obtained scattering law is based on the solution of the general kinematic problem of elastic scattering of neutrons on nuclei in the L-system. Both the neutron and the nucleus possess arbitrary velocities in the L-system. For the new scattering law we obtain the flux densities and neutron moderation spectra as functions of temperature for the reactor fissile medium. The expressions for the moderating neutrons spectra allow reinterpreting the physical nature of the underlying processes in the thermal region. (orig.)

  7. Transmutation of Minor Actinide in well thermalized neutron field and application of advanced neutron source (ANS)

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Hirakawa, Naohiro

    1995-01-01

    Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246 Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246 Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100 eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10 -5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation. (author)

  8. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  9. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  12. Neutron Thermalization and Reactor Spectra. Vol. II. Proceedings of the Symposium on Neutron Thermalization and Reactor Spectra

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held at Ann Arbor, Michigan, USA, 17 - 21 July 1967. The meeting was attended by 143 participants from 24 Member States and one international organization. Contents: (Vol.I) Theory of neutron thermalization (15 papers); Scattering law (20 papers); Angular, space, temperature and time dependence of neutron spectra (9 papers). (Vol.II) Measurement of thermal neutron spectra and spectral indices, and comparison with theory (17 papers); Time-dependent problems in neutron thermalization (12 papers). Each paper is in its original language (61 English, 1 French and 11 Russian) and is preceded by an abstract in English with one in the original language if this is not English. Discussions are in English.

  13. On the use of silicon as thermal neutron filter

    International Nuclear Information System (INIS)

    Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

    2003-01-01

    A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 μeV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

  14. On the use of silicon as thermal neutron filter

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Habib, N.; Ashry, A.; Fathalla, M. E-mail: mohamedfathalla@hotmail.com

    2003-12-01

    A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 {mu}eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.

  15. Chemical warfare agents identification by thermal neutron detection

    International Nuclear Information System (INIS)

    Liu Boxue; Ai Xianyun; Tan Daoyuan; Zhang Dianqin

    2000-01-01

    The hydrogen concentration determination by thermal neutron detection is a non-destructive, fast and effective method to identify chemical warfare agents and TNT that contain different hydrogen fraction. When an isotropic neutron source is used to irradiate chemical ammunition, hydrogen atoms of the agent inside shell act as a moderator and slow down neutrons. The number of induced thermal neutrons depends mainly upon hydrogen content of the agent. Therefore measurement of thermal neutron influence can be used to determine hydrogen atom concentration, thereby to determine the chemical warfare agents. Under a certain geometry three calibration curves of count rate against hydrogen concentration were measured. According to the calibration curves, response of a chemical agent or TNT could be calculated. Differences of count rate among chemical agents and TNT for each kind of shells is greater than five times of standard deviations of count rate for any agent, so chemical agents or TNT could be identified correctly. Meanwhile, blast tube or liquid level of chemical warfare agent could affect the response of thermal neutron count rate, and thereby the result of identification. (author)

  16. Impact of neutron irradiation on thermal helium desorption from iron

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

    2017-06-15

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  17. Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

    Science.gov (United States)

    Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

    2018-01-01

    Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

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

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

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

  19. Thermal compression modulus of polarized neutron matter

    International Nuclear Information System (INIS)

    Abd-Alla, M.

    1990-05-01

    We applied the equation of state for pure polarized neutron matter at finite temperature, calculated previously, to calculate the compression modulus. The compression modulus of pure neutron matter at zero temperature is very large and reflects the stiffness of the equation of state. It has a little temperature dependence. Introducing the spin excess parameter in the equation of state calculations is important because it has a significant effect on the compression modulus. (author). 25 refs, 2 tabs

  20. Applications for coupled core neutronics and thermal-hydraulic models

    International Nuclear Information System (INIS)

    Eller, J.

    1996-01-01

    The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

  1. Boron neutron capture therapy (BNCT). Recent aspect, a change from thermal neutron to epithermal neutron beam and a new protocol

    International Nuclear Information System (INIS)

    Nakagawa, Yoshinobu

    1999-01-01

    Since 1968, One-hundred seventy three patients with glioblastoma (n=81), anaplastic astrocytoma (n=44), low grade astrocytoma (n=16) or other types of tumor (n=32) were treated by boron-neutron capture therapy (BNCT) using a combination of thermal neutron and BSH in 5 reactors (HTR n=13, JRR-3 n=1, MuITR n=98, KUR n=28, JRR-2 n=33). Out of 101 patients with glioma treated by BNCT under the recent protocol, 33 (10 glioblastoma, 14 anaplastic astrocytoma, 9 low grade astrocytoma) patients lived or have lived longer than 3 years. Nine of these 33 lived or have lived longer than 10 years. According to the retrospective analysis, the important factors related to the clinical results were tumor dose radiation dose and maximum radiation dose in thermal brain cortex. The result was not satisfied as it was expected. Then, we decided to introduce mixed beams which contain thermal neutron and epithermal neutron beams. KUR was reconstructed in 1996 and developed to be available to use mixed beams. Following the shutdown of the JRR-2, JRR-4 was renewed for medical use in 1998. Both reactors have capacity to yield thermal neutron beam, epithermal neutron beam and mixed beams. The development of the neutron source lead us to make a new protocol. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  3. Thermal neutron scattering studies of condensed matter under high pressures

    International Nuclear Information System (INIS)

    Carlile, C.J.; Salter, D.C.

    1978-01-01

    Although temperature has been used as a thermodynamic variable for samples in thermal neutron scattering experiments since the inception of the neutron technique, it is only in the last decade that high pressures have been utilised for this purpose. In the paper the problems particular to this field of work are outlined and a review is made of the types of high-pressure cells used and the scientific results obtained from the experiments. 103 references. (author)

  4. Determination of thermal neutrons diffusion length in graphite

    International Nuclear Information System (INIS)

    Garcia Fite, J.

    1959-01-01

    The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs

  5. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Directory of Open Access Journals (Sweden)

    Hu J.-P.

    2016-01-01

    Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

  6. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

    2014-05-23

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

  7. Method and apparatus for measuring thermal neutron characteristics

    International Nuclear Information System (INIS)

    Johnstone, C.W.

    1983-01-01

    The thermal neutron decay characteristics of an earth formation are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. These measurement intervals may comprise a sequence of time gates following a delay after the neutron burst. The duration of the neutron bursts, of the delay between the burst and the start of the sequence, and of the individual time gates, may all be adjusted by a common, selected one of a finite number of scale factor values. The set of two measurement intervals is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and a function of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. (author)

  8. Introduction to the theory of thermal neutron scattering

    CERN Document Server

    Squires, G L

    2012-01-01

    Since the advent of the nuclear reactor, thermal neutron scattering has proved a valuable tool for studying many properties of solids and liquids, and research workers are active in the field at reactor centres and universities throughout the world. This classic text provides the basic quantum theory of thermal neutron scattering and applies the concepts to scattering by crystals, liquids and magnetic systems. Other topics discussed are the relation of the scattering to correlation functions in the scattering system, the dynamical theory of scattering and polarisation analysis. No previous knowledge of the theory of thermal neutron scattering is assumed, but basic knowledge of quantum mechanics and solid state physics is required. The book is intended for experimenters rather than theoreticians, and the discussion is kept as informal as possible. A number of examples, with worked solutions, are included as an aid to the understanding of the text.

  9. Targets for bulk hydrogen analysis using thermal neutrons

    CERN Document Server

    Csikai, J; Buczko, C M

    2002-01-01

    The reflection property of substances can be characterized by the reflection cross-section of thermal neutrons, sigma subbeta. A combination of the targets with thin polyethylene foils allowed an estimation of the flux depression of thermal neutrons caused by a bulk sample containing highly absorbing elements or compounds. Some new and more accurate sigma subbeta values were determined by using the combined target arrangement. For the ratio, R of the reflection and the elastic scattering cross-sections of thermal neutrons, R=sigma subbeta/sigma sub E sub L a value of 0.60+-0.02 was found on the basis of the data obtained for a number of elements from H to Pb. Using this correlation factor, and the sigma sub E sub L values, the unknown sigma subbeta data can be deduced. The equivalent thicknesses, to polyethylene or hydrogen, of the different target materials were determined from the sigma subbeta values.

  10. Thermal neutron dose calculation in synovium membrane for BNCS

    International Nuclear Information System (INIS)

    Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.

    2006-01-01

    A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values. (author)

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

    International Nuclear Information System (INIS)

    Thiagu Supramaniam

    2007-01-01

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

  12. Flux-Vortex Pinning and Neutron Star Evolution

    Indian Academy of Sciences (India)

    M. Ali Alpar

    2017-09-12

    Sep 12, 2017 ... M. ALI ALPAR. Faculty of Engineering and Natural Sciences, Sabancı University, 34956, Istanbul, Turkey. E-mail: ... netic field of the neutron star were B ∼ 109 G. At the ..... across pinning energy barriers by thermal activation.

  13. Status of thermal neutron scattering data for graphite

    International Nuclear Information System (INIS)

    Mattes, M.; Keinert, J.

    2005-07-01

    At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross sections and the angular and energy distributions of the scattered neutrons. These effects are described in the thermal sub-library of evaluated files in File 7 of the ENDF-6 format. A re-evaluation of thermal neutron scattering data for carbon bound in graphite has been performed to investigate the impact of models (e.g., generalised frequency distributions) based on different experimental and theoretical data for the generation of scattering law data files S(α,β,T) and coherent elastic scattering data. Two phonon frequency distributions of graphite published in 2002 and 2004 were considered and the results compared with those based on the phonon spectra from Koppel et al. (published in 1968), on which the evaluations of ENDF/B-VI and JEFF-3.1 are based. The new frequency distributions were partly derived from ab initio simulations. Detailed comparisons with measurements of differential and integral neutron cross sections and other relevant data are reported. In addition, thermal MCNP data sets for use in the continuous Monte Carlo codes MCNP and MCNPX were generated from these evaluations for different temperatures. Calculated neutron spectra were found to be in good agreement with the measurements. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

  15. Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

    International Nuclear Information System (INIS)

    Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

    1994-01-01

    Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

  16. Neutron star evolution and the structure of matter at high density

    International Nuclear Information System (INIS)

    Soyeur, Madeleine.

    1981-09-01

    The structure and properties of neutron stars are determined by the state of cold nuclear matter at high density. In order to investigate the behavior of matter inside neutron stars, observables sensitive to their internal structure have to be calculated and confronted to observations. The thermal radiation of neutron stars seems to be a good candidate to be such observable. It can be shown that the neutrino luminosity of neutron stars, responsible for their cooling in the early stages of their evolution is strongly dependent on possible phase transitions to superfluid nucleons, to pion condensation or to quark matter. The specific heat of matter is also not the same in the various phases expected at high density and is particularly sensitive to the nucleon superfluidity. At present, both the theoretical estimates and the observations of the thermal properties of neutron stars are still quite preliminary. In particular, large uncertainties due to possible reheating mechanisms and magnetic field effects make the theoretical interpretation of the steady radiation of pulsars quite difficult

  17. Transmission of Thermal Neutrons through Boral

    Energy Technology Data Exchange (ETDEWEB)

    Aakerhielm, F

    1960-06-15

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

  18. Transmission of Thermal Neutrons through Boral

    International Nuclear Information System (INIS)

    Aakerhielm, F.

    1960-06-01

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

  19. Qualification of the neutronic evolution of LWR fuels in MELUSINE

    International Nuclear Information System (INIS)

    Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.

    1984-09-01

    MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr

  20. Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

    Energy Technology Data Exchange (ETDEWEB)

    Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

    2002-07-01

    Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

  1. The thermal evolution of universe: standard model

    International Nuclear Information System (INIS)

    Nascimento, L.C.S. do.

    1975-08-01

    A description of the dynamical evolution of the Universe following a model based on the theory of General Relativity is made. The model admits the Cosmological principle,the principle of Equivalence and the Robertson-Walker metric (of which an original derivation is presented). In this model, the universe is considered as a perfect fluid, ideal and symmetric relatively to the number of particles and antiparticles. The thermodynamic relations deriving from these hypothesis are derived, and from them the several eras of the thermal evolution of the universe are established. Finally, the problems arising from certain specific predictions of the model are studied, and the predictions of the abundances of the elements according to nucleosynthesis and the actual behavior of the universe are analysed in detail. (author) [pt

  2. Earth evolution as a thermal system

    Science.gov (United States)

    Tang, C.

    2014-12-01

    After fifty years of plate-tectonic theory, the reasons why earth sometime freezed as a snowball or sometime became lethally hot resulting in mass extinction remain enigmatic. This article proposes a new hypothesis on Earth evolution. The unbalance of heat between the input and output is considered as the driving force for the Earth evolution, the lithospheric expansion and associated uplift are the triggers, the self-organized progressive failure leading to collapse of the Earth are the amplifier, and the global scale response in terms of volcanism and magmatism is the globalizer. This shallow process of lithosphere may reach a critical state with a positive feedback loop, and result in the formation of no-plume original Large Igneous Provinces (NPOLIP) in a top-down pattern. Endothermic phase changes during de-compressive melting remove heat from and cool their surroundings, including the upper parts of the lithosphere. The huge loss of Earth's heat during eruption of LIPs, together with the endothermic cooling, may put the thermal cycle to an end and a new start of the cycle initiates. In summary, Earth drives itself to evolve in terms of thermal cycles. Global cooling and warming are the two stages of the many cycles during the Earth evolution. Glaciations are the extreme result of global cooling, whereas the LIPs, sometime accompanied with remarkable sea level dropping, are the extreme result of global warming, with a long recovering age, the interglacialstage, between them. They come and go as thermal cycle evolves, with climate warming, being caused by Earth itself rather than by external forces or human activities, as the most attractive prediction.

  3. The early thermal evolution of Mars

    Science.gov (United States)

    Bhatia, G. K.; Sahijpal, S.

    2016-01-01

    Hf-W isotopic systematics of Martian meteorites have provided evidence for the early accretion and rapid core formation of Mars. We present the results of numerical simulations performed to study the early thermal evolution and planetary scale differentiation of Mars. The simulations are confined to the initial 50 Myr (Ma) of the formation of solar system. The accretion energy produced during the growth of Mars and the decay energy due to the short-lived radio-nuclides 26Al, 60Fe, and the long-lived nuclides, 40K, 235U, 238U, and 232Th are incorporated as the heat sources for the thermal evolution of Mars. During the core-mantle differentiation of Mars, the molten metallic blobs were numerically moved using Stoke's law toward the center with descent velocity that depends on the local acceleration due to gravity. Apart from the accretion and the radioactive heat energies, the gravitational energy produced during the differentiation of Mars and the associated heat transfer is also parametrically incorporated in the present work to make an assessment of its contribution to the early thermal evolution of Mars. We conclude that the accretion energy alone cannot produce widespread melting and differentiation of Mars even with an efficient consumption of the accretion energy. This makes 26Al the prime source for the heating and planetary scale differentiation of Mars. We demonstrate a rapid accretion and core-mantle differentiation of Mars within the initial ~1.5 Myr. This is consistent with the chronological records of Martian meteorites.

  4. Thermal neutron calibration channel at LNMRI/IRD

    International Nuclear Information System (INIS)

    Astuto, A.; Salgado, A.P.; Lopes, R.T.; Leite, S.P.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

    2014-01-01

    The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four 241 Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units. The pile construction form using blocks allows distinct arrangements for new studies and possibilities of other LNMRI reference fields. The results can be predicted by the simulation used in this work. Different number of each type of blocks and sources can be used. The main difference observed between the final measurement and simulation results might be due to the difference in composition of paraffin blocks used in assembling the pile. In order to confirm the thermal neutron field and fluence rate in the central chamber (inside the channel) that will be used to irradiate small neutron detectors, it is necessary to use another quantification method such as the gold foils activation with measurement traceability. It will be performed in a future stage. (authors)

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

    Science.gov (United States)

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

    2017-05-16

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

  6. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

    Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

    2005-01-01

    Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

  7. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hee Joon [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Edwards, Dan J., E-mail: dan.edwards@pnnl.gov [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Kurtz, Richard J. [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Yamamoto, Takuya; Wu, Yuan [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Odette, G. Robert [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Materials Department, University of California, Santa Barbara, CA 93106 (United States)

    2017-02-15

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron {sup 58}Ni(n{sub th},γ) {sup 59}Ni(n{sub th},α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al{sub 2}YO{sub 3} oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  8. Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

  9. Microscopic cross-section measurements by thermal neutron activation

    International Nuclear Information System (INIS)

    Avila L, J.

    1987-08-01

    Microscopic cross sections measured by thermal neutron activation using RP-0 reactor at the Peruvian Nuclear Energy Institute. The method consists in measuring microscopic cross section ratios through activated samples, requiring being corrected in thermal and epithermal energetic range by Westcott formalism. Furthermore, the comptage ratios measured for each photopeak to its decay fraction should be normalized from interrelation between both processes above, activation microscopic cross sections are obtained

  10. Thermal hydraulic and neutronic interaction in the rotating bed reactor

    International Nuclear Information System (INIS)

    Lee, C.C.

    1986-01-01

    Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

  11. Specimen environments in thermal neutron scattering experiments

    International Nuclear Information System (INIS)

    Cebula, D.J.

    1980-11-01

    This report is an attempt to collect into one place outline information concerning the techniques used and basic design of sample environment apparatus employed in neutron scattering experiments. Preliminary recommendations for the specimen environment programme of the SNS are presented. The general conclusion reached is that effort should be devoted towards improving reliability and efficiency of operation of specimen environment apparatus and developing systems which are robust and easy to use, rather than achieving performance at the limits of technology. (author)

  12. Set of thermal neutron-scattering experiments for the Weapons Neutron Research Facility

    International Nuclear Information System (INIS)

    Brugger, R.M.

    1975-12-01

    Six classes of experiments form the base of a program of thermal neutron scattering at the Weapons Neutron Research (WNR) Facility. Three classes are to determine the average microscopic positions of atoms in materials and three are to determine the microscopic vibrations of these atoms. The first three classes concern (a) powder sample neutron diffraction, (b) small angle scattering, and (c) single crystal Laue diffraction. The second three concern (d) small kappa inelastic scattering, (e) scattering surface phonon measurements, and (f) line widths. An instrument to couple with the WNR pulsed source is briefly outlined for each experiment

  13. Thermal Neutron Capture and Thermal Neutron Burn-up of K isomeric state of 177mLu: a way to the Neutron Super-Elastic Scattering cross section

    International Nuclear Information System (INIS)

    Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.

    2006-01-01

    Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron

  14. Need for nuclear data for thermal neutron reactors

    International Nuclear Information System (INIS)

    Bouchard, J.; Golinelli, C.; Tellier, H.

    1983-01-01

    The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

  15. Measurement of the diffusion length of thermal neutrons inside graphite

    International Nuclear Information System (INIS)

    Ertaud, A.; Beauge, R.; Fauquez, H.; De Laboulay, H.; Mercier, C.; Vautrey, L.

    1948-11-01

    The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra α → Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm ± 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

  16. High temperature ductility of austenitic alloys exposed to thermal neutrons

    International Nuclear Information System (INIS)

    Watanabe, K.; Kondo, T.; Ogawa, Y.

    1982-01-01

    Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

  17. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  18. Parallel diffusion length on thermal neutrons in rod type lattices

    International Nuclear Information System (INIS)

    Ahmed, T.; Siddiqui, S.A.M.M.; Khan, A.M.

    1981-11-01

    Calculation of diffusion lengths of thermal neutrons in lead-water and aluminum water lattices in direction parallel to the rods are performed using one group diffusion equation together with Shevelev transport correction. The formalism is then applied to two practical cases, the Kawasaki (Hitachi) and the Douglas point (Candu) reactor lattices. Our results are in good agreement with the observed values. (author)

  19. 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

  20. A two-medium thermal neutron spectrum programme

    International Nuclear Information System (INIS)

    Bindon, D.C.

    1960-07-01

    A computer programme is described for computing the thermal neutron spectra and effective cross-sections in a reactor system of two media by the method of H. Takahashi. The programme has been prepared and tested for use with the Ferranti Mercury computer. (author)

  1. Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

    Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

  2. Real time neutronic evolution CNE (Embalse nuclear power plant)

    International Nuclear Information System (INIS)

    Notari, C.; Waldman, R.M.

    1993-01-01

    The simulator of the Embalse nuclear power plant uses a Point Reactor Model(PRM) for the neutronic evolution calculation. As this model is not conservative for transients produced by the sudden or localized reactivity insertion in big cores, it is convenient to use spatial models in these cases. In this report we show the results obtained using a nodal model (codes NODOS-TIEMPO). This model has been fitted against a more exact solution for the neutron flux and delayed neutron precursors. This has been done for the reactor at full power with nominal values for the reactivity control devices (liquid zones and adjusters rods). Transients corresponding to the global variation of the liquid zones and to the insertion of fresh fuel in some channels are shown. The results are compared with calculations made with the quasi-static model of the PUMA code. (author). 1 ref

  3. Thermal expansion study of simulated DUPIC fuel using neutron diffraction

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Ryu, H. J.; Bae, J. H.; Kim, H. S.; Song, K. C.; Yang, M. S.; Choi, Y. N.; Han, Y. S.; Oh, H. S.

    2001-07-01

    The lattice parameters of simulated DUPIC fuel and UO2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO2 and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO2. For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO2 and simulated DUPIC fuel are 10.471 ''10-6 and 10.751 ''10-6 K-1, respectively

  4. Detection of land mines using fast and thermal neutron analysis

    International Nuclear Information System (INIS)

    Bach, P.

    1998-01-01

    The detection of land mines is made possible by using nuclear sensor based on neutron interrogation. Neutron interrogation allows to detect the sensitive elements (C, H, O, N) of the explosives in land mines or in unexploded shells: the evaluation of characteristic ratio N/O and C/O in a volume element gives a signature of high explosives. Fast neutron interrogation has been qualified in our laboratories as a powerful close distance method for identifying the presence of a mine or explosive. This method could be implemented together with a multisensor detection system - for instance IR or microwave - to reduce the false alarm rate by addressing the suspected area. Principle of operation is based on the measurement of gamma rays induced by neutron interaction with irradiated nuclei from the soil and from a possible mine. Specific energy of these gamma rays allows to recognise the elements at the origin of neutron interaction. Several detection methods can be used, depending on nuclei to be identified. Analysis of physical data, computations by simulation codes, and experimentations performed in our laboratory have shown the interest of Fast Neutron Analysis (FNA) combined with Thermal Neutron Analysis (TNA) techniques, especially for detection of nitrogen 14 N, carbon 12 C and oxygen 16 O. The FNA technique can be implemented using a 14 MeV sealed neutron tube, and a set of detectors. The mines detection has been demonstrated from our investigations, using a low power neutron generator working in the 10 8 n/s range, which is reasonable when considering safety rules. A fieldable demonstrator would be made with a detection head including tube and detectors, and with remote electronics, power supplies and computer installed in a vehicle. (author)

  5. Boron thermal/epithermal neutron capture therapy

    International Nuclear Information System (INIS)

    Fairchild, R.G.

    1982-01-01

    The development of various particle beams for radiotherapy represents an attempt to improve dose distribution, and to provide high LET radiations which are less sensitive to ambient physical and radiobiological factors such as oxygen tension, cell cycle, and dose rate. In general, a compromise is necessary as effective RBE is reduced in order to spread the dose distribution over the anticipated tumor volume. The approach of delivering stable non-toxic isotopes to tumor, and then activating these atoms subsequently via an external radiation beam has mator advantages; problems associated with high uptake of these isotopes in competing cell pools are obviated, and the general tumor volume can be included in the treatment field of the activating beam. As long as the normal tissues supporting tumor show a low uptake of the isotope to be activated, and as long as the range of the reaction products is short, dose will be restricted to tumor, with a consequent high therapeutic ratio. Neutron Capture Therapy (NCT) is generally carried out by activating boron-10 with low energy neutrons. The range of the high LET, low OER particles from the 10 B(n, α) 7 Li reaction is approx. 10μ, or one cell diameter, a situation that is optimal for cell killing. Significant advantages may be gained by using the NCT procedure in conjunction with improved tissue penetration provided with epithermal or filtered beams, and new compounds showing physiological binding to tumor

  6. Effects of high thermal neutron fluences on Type 6061 aluminum

    International Nuclear Information System (INIS)

    Weeks, J.R.; Czajkowski, C.J.; Farrell, K.

    1992-01-01

    The control rod drive follower tubes of the High Flux Beam Reactor are contructed from precipitation-hardened 6061-T6 aluminum alloy and they operate in the high thermal neutron flux regions of the core. It is shown that large thermal neutron fluences up to ∼4 x 10 23 n/cm 2 at 333K cause large increases in tensile strength and relatively modest decreases in tensile elongation while significantly reducing the notch impact toughness at room temperature. These changes are attributed to the development of a fine distribution of precipitates of amorphous silicon of which about 8% is produced radiogenically. A proposed role of thermal-to-fast flux ratio is discussed

  7. Scattering of thermal neutron by the water molecule

    International Nuclear Information System (INIS)

    Rosa, L.P.

    The calculation of the differenctial cross section for scattering of thermal neutrons by water, taking into account the translational, rotational and vibrational motions of the water molecule, is presented according to Nelkin' model. Some modifications are presented which have been introduced in the original method to improve the results and an application has been made to reactor physics, by calculating the thermal neutron flux in a homogenous medium containing water and absorver. Thirty thermal energy groups have been used to compute the spectra. Within the limits of error, better agreement has been obtained between theory and experiments by using a modified Nelkin kernel consisting of substituting the asymptotic formulae for the rotational and vibrational motions by more exact expressions, similar to the Buttler model for heavy water

  8. Thermal and Chemical Evolution of Collapsing Filaments

    Energy Technology Data Exchange (ETDEWEB)

    Gray, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Scannapieco, Evan [Arizona State Univ., Mesa, AZ (United States). School of Earth and Space Exploration

    2013-01-15

    Intergalactic filaments form the foundation of the cosmic web that connect galaxies together, and provide an important reservoir of gas for galaxy growth and accretion. Here we present very high resolution two-dimensional simulations of the thermal and chemical evolution of such filaments, making use of a 32 species chemistry network that tracks the evolution of key molecules formed from hydrogen, oxygen, and carbon. We study the evolution of filaments over a wide range of parameters including the initial density, initial temperature, strength of the dissociating UV background, and metallicity. In low-redshift, Z ≈ 0.1Z filaments, the evolution is determined completely by the initial cooling time. If this is sufficiently short, the center of the filament always collapses to form dense, cold core containing a substantial fraction of molecules. In high-redshift, Z = 10-3Z filaments, the collapse proceeds much more slowly. This is due mostly to the lower initial temperatures, which leads to a much more modest increase in density before the atomic cooling limit is reached, making subsequent molecular cooling much less efficient. Finally, we study how the gravitational potential from a nearby dwarf galaxy affects the collapse of the filament and compare this to NGC 5253, a nearby starbusting dwarf galaxy thought to be fueled by the accretion of filament gas. In contrast to our fiducial case, a substantial density peak forms at the center of the potential. This peak evolves faster than the rest of the filament due to the increased rate at which chemical species form and cooling occur. We find that we achieve similar accretion rates as NGC 5253, but our two-dimensional simulations do not recover the formation of the giant molecular clouds that are seen in radio observations.

  9. Experimental investigation of thermal neutron analysis based landmine detection technology

    International Nuclear Information System (INIS)

    Zeng Jun; Chu Chengsheng; Ding Ge; Xiang Qingpei; Hao Fanhua; Luo Xiaobing

    2013-01-01

    Background: Recently, the prompt gamma-rays neutron activation analysis method is wildly used in coal analysis and explosive detection, however there were less application about landmine detection using neutron method especially in the domestic research. Purpose: In order to verify the feasibility of Thermal Neutron Analysis (TNA) method used in landmine detection, and explore the characteristic of this technology. Methods: An experimental system of TNA landmine detection was built based on LaBr 3 (Ce) fast scintillator detector and 252 Cf isotope neutron source. The system is comprised of the thermal neutron transition system, the shield system, and the detector system. Results: On the basis of the TNA, the wide energy area calibration method especially to the high energy area was investigated, and the least detection time for a typical mine was defined. In this study, the 72-type anti-tank mine, the 500 g TNT sample and several interferential objects are tested in loess, red soil, magnetic soil and sand respectively. Conclusions: The experimental results indicate that TNA is a reliable demining method, and it can be used to confirm the existence of Anti-Tank Mines (ATM) and large Anti-Personnel Mines (APM) in complicated condition. (authors)

  10. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  11. Thermal evolution of the crystal structure of the orthorhombic perovskite LaFeO3

    International Nuclear Information System (INIS)

    Dixon, Charlotte A.L.; Kavanagh, Christopher M.; Knight, Kevin S.; Kockelmann, Winfried; Morrison, Finlay D.; Lightfoot, Philip

    2015-01-01

    The thermal evolution of the crystal structure of the prototypical orthorhombic perovskite LaFeO 3 has been studied in detail by powder neutron diffraction in the temperature range 25thermal behavior to be understood. In particular, the largest-amplitude symmetry modes (viz. in-phase and out-of-phase octahedral tilts, and A-site cation displacements) are shown to display relatively ‘normal’ behavior, increasing with decreasing temperature, which contrasts with the anomalous behavior previously shown by the derivative Bi 0.5 La 0.5 FeO 3 . However, an unexpected behavior is seen in the nature of the intra-octahedral distortion, which is used to rationalize the unique occurrence of a temperature dependent crossover of the a and c unit cell metrics in this compound. - Graphical abstract: The unusual thermal evolution of lattice metrics in the perovskite LaFeO 3 is rationalized from a detailed powder neutron diffraction study. - Highlights: • Crystal structure of the perovskite LaFeO 3 studied in detail by powder neutron diffraction. • Unusual thermal evolution of lattice metrics rationalized. • Contrasting behavior to Bi-doped LaFeO 3 . • Octahedral distortion/tilt parameters explain unusual a and c lattice parameter behavior

  12. New thermal neutron calibration channel at LNMRI/IRD

    Energy Technology Data Exchange (ETDEWEB)

    Astuto, A.; Lopes, R.T., E-mail: achillesbr@gmail.com [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ/LNMRI), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiacoes Ionizantes

    2015-07-01

    A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four {sup 241}Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

  13. New thermal neutron calibration channel at LNMRI/IRD

    International Nuclear Information System (INIS)

    Astuto, A.; Lopes, R.T.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

    2015-01-01

    A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241 Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

  14. Thermal neutron absorption cross section of small samples

    International Nuclear Information System (INIS)

    Nghiep, T.D.; Vinh, T.T.; Son, N.N.; Vuong, T.V.; Hung, N.T.

    1989-01-01

    A modified steady method for determining the macroscopic thermal neutron absorption cross section of small samples 500 cm 3 in volume is described. The method uses a moderating block of paraffin, Pu-Be neutron source emitting 1.1x10 6 n.s. -1 , SNM-14 counter and ordinary counting equipment. The interval of cross section from 2.6 to 1.3x10 4 (10 -3 cm 2 g -1 ) was measured. The experimental data are described by calculation formulae. 7 refs.; 4 figs

  15. Thermal neutron scattering cross sections of beryllium and magnesium oxides

    International Nuclear Information System (INIS)

    Al-Qasir, Iyad; Jisrawi, Najeh; Gillette, Victor; Qteish, Abdallah

    2016-01-01

    Highlights: • Neutron thermalization in BeO and MgO was studied using Ab initio lattice dynamics. • The BeO phonon density of states used to generate the current ENDF library has issues. • The BeO cross sections can provide a more accurate ENDF library than the current one. • For MgO an ENDF library is lacking: a new accurate one can be built from our results. • BeO is a better filter than MgO, especially when cooled down to 77 K. - Abstract: Alkaline-earth beryllium and magnesium oxides are fundamental materials in nuclear industry and thermal neutron scattering applications. The calculation of the thermal neutron scattering cross sections requires a detailed knowledge of the lattice dynamics of the scattering medium. The vibrational properties of BeO and MgO are studied using first-principles calculations within the frame work of the density functional perturbation theory. Excellent agreement between the calculated phonon dispersion relations and the experimental data have been obtained. The phonon densities of states are utilized to calculate the scattering laws using the incoherent approximation. For BeO, there are concerns about the accuracy of the phonon density of states used to generate the current ENDF/B-VII.1 libraries. These concerns are identified, and their influences on the scattering law and inelastic scattering cross section are analyzed. For MgO, no up to date thermal neutron scattering cross section ENDF library is available, and our results represent a potential one for use in different applications. Moreover, the BeO and MgO efficiencies as neutron filters at different temperatures are investigated. BeO is found to be a better filter than MgO, especially when cooled down, and cooling MgO below 77 K does not significantly improve the filter’s efficiency.

  16. Computed tomography with thermal neutrons and gaseous position sensitive detector

    International Nuclear Information System (INIS)

    Souza, Maria Ines Silvani

    2001-12-01

    A third generation tomographic system using a parallel thermal neutron beam and gaseous position sensitive detector has been developed along three discrete phases. At the first one, X-ray tomographic images of several objects, using a position sensitive detector designed and constructed for this purpose have been obtained. The second phase involved the conversion of that detector for thermal neutron detection, by using materials capable to convert neutrons into detectable charged particles, testing afterwards its performance in a tomographic system by evaluation the quality of the image arising from several test-objects containing materials applicable in the engineering field. High enriched 3 He, replacing the argon-methane otherwise used as filling gas for the X-ray detection, as well as, a gadolinium foil, have been utilized as converters. Besides the pure enriched 3 He, its mixture with argon-methane and later on with propane, have been also tested, in order to evaluate the detector efficiency and resolution. After each gas change, the overall performance of the tomographic system using the modified detector, has been analyzed through measurements of the related parameters. This was done by analyzing the images produced by test-objects containing several materials having well known attenuation coefficients for both thermal neutrons and X-rays. In order to compare the performance of the position sensitive detector as modified to detect thermal neutrons, with that of a conventional BF 3 detector, additional tomographs have been conducted using the last one. The results have been compared in terms of advantages, handicaps and complementary aspects for different kinds of radiation and materials. (author)

  17. Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

    When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

  18. Analytical modeling of thin film neutron converters and its application to thermal neutron gas detectors

    Energy Technology Data Exchange (ETDEWEB)

    Piscitelli, F; Esch, P Van, E-mail: piscitelli@ill.fr [Institut Laue-Langevin (ILL), 6, Jules Horowitz, 38042 Grenoble (France)

    2013-04-15

    A simple model is explored mainly analytically to calculate and understand the PHS of single and multi-layer thermal neutron detectors and to help optimize the design in different circumstances. Several theorems are deduced that can help guide the design.

  19. Thermal force approach to molecular evolution.

    Science.gov (United States)

    Braun, Dieter; Libchaber, Albert

    2004-06-01

    Recent experiments are discussed where temperature gradients across mesoscopic pores are shown to provide essential mechanisms for autonomous molecular evolution. On the one hand, laminar thermal convection can drive DNA replication as the molecules are continuously cycled between hot and cold regions of a chamber. On the other hand, thermophoresis can accumulate charged biopolymers in similar convection settings. The experiments show that temperature differences analogous to those across porous rocks present a robust nonequilibrium boundary condition to feed the replication and accumulation of evolving molecules. It is speculated that similar nonequilibrium conditions near porous submarine hydrothermal mounds could have triggered the origin of life. In such a scenario, the encapsulation of cells with membranes would be a later development. It is expected that detailed studies of mesoscopic boundary conditions under nonequilibrium conditions will reveal new connecting pieces in the fascinating puzzle of the origins of life.

  20. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

    Science.gov (United States)

    Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

    2011-12-01

    A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed

  1. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

    2011-12-15

    Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and

  2. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

    International Nuclear Information System (INIS)

    Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo

    2011-01-01

    Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global

  3. Thermal neutron scattering kernels for sapphire and silicon single crystals

    International Nuclear Information System (INIS)

    Cantargi, F.; Granada, J.R.; Mayer, R.E.

    2015-01-01

    Highlights: • Thermal cross section libraries for sapphire and silicon single crystals were generated. • Debye model was used to represent the vibrational frequency spectra to feed the NJOY code. • Sapphire total cross section was measured at Centro Atómico Bariloche. • Cross section libraries were validated with experimental data available. - Abstract: Sapphire and silicon are materials usually employed as filters in facilities with thermal neutron beams. Due to the lack of the corresponding thermal cross section libraries for those materials, necessary in calculations performed in order to optimize beams for specific applications, here we present the generation of new thermal neutron scattering kernels for those materials. The Debye model was used in both cases to represent the vibrational frequency spectra required to feed the NJOY nuclear data processing system in order to produce the corresponding libraries in ENDF and ACE format. These libraries were validated with available experimental data, some from the literature and others obtained at the pulsed neutron source at Centro Atómico Bariloche

  4. Study of boron carbide evolution under neutron irradiation

    International Nuclear Information System (INIS)

    Simeone, D.

    1999-01-01

    Owing to its high neutron efficiency, boron carbide (B 4 C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the 10 B(n,α) 7 Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 4 C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 4 C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B 4 C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B 4 C. (author)

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

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

  6. Physics of epi-thermal boron neutron capture therapy (epi-thermal BNCT).

    Science.gov (United States)

    Seki, Ryoichi; Wakisaka, Yushi; Morimoto, Nami; Takashina, Masaaki; Koizumi, Masahiko; Toki, Hiroshi; Fukuda, Mitsuhiro

    2017-12-01

    The physics of epi-thermal neutrons in the human body is discussed in the effort to clarify the nature of the unique radiologic properties of boron neutron capture therapy (BNCT). This discussion leads to the computational method of Monte Carlo simulation in BNCT. The method is discussed through two examples based on model phantoms. The physics is kept at an introductory level in the discussion in this tutorial review.

  7. Study of thermal neutron capture in 58 Ni

    International Nuclear Information System (INIS)

    Carbonari, A.W.; Pecequilo, B.R.S.

    1988-08-01

    The energies and intensities of the primary gamma-rays from 58 Ni (n, γ) 59 Ni reaction have been measured with a Ge(li) pair-spectrometer in the region of 3.7 to 9.3 MeV. The thermal neutron capture cross section of 58 Ni was determined to be 4.52 +- 0.10 by summing the primary transition intensities. The neutron separation energy was found to be 8999.93 +- 0.34 KeV. It is shown that the decay of the capture state is non-statistical and that there is a strong correlation between the strengths of excitation of levels by the (n, γ) and (d,p) reactions. These results are discussed in terms of a direct neutron capture reaction mechanism. (author) [pt

  8. Methods and apparatus for environmental correction of thermal neutron logs

    International Nuclear Information System (INIS)

    Preeg, W.E.; Scott, H.D.

    1983-01-01

    An on-line environmentally-corrected measurement of the thermal neutron decay time (tau) of an earth formation traversed by a borehole is provided in a two-detector, pulsed neutron logging tool, by measuring tau at each detector and combining the two tau measurements in accordance with a previously established empirical relationship of the general form: tau = tausub(F) +A(tausub(F) + tausub(N)B) + C, where tausub(F) and tausub(N) are the tau measurements at the far-spaced and near-spaced detectors, respectively, A is a correction coefficient for borehole capture cross section effects, B is a correction coefficient for neutron diffusion effects, and C is a constant related to parameters of the logging tool. Preferred numerical values of A, B and C are disclosed, and a relationship for more accurately approximating the A term to specific borehole conditions. (author)

  9. Thermal diffuse scattering in angular-dispersive neutron diffraction

    International Nuclear Information System (INIS)

    Popa, N.C.; Willis, B.T.M.

    1998-01-01

    The theoretical treatment of one-phonon thermal diffuse scattering (TDS) in single-crystal neutron diffraction at fixed incident wavelength is reanalysed in the light of the analysis given by Popa and Willis [Acta Cryst. (1994), (1997)] for the time-of-flight method. Isotropic propagation of sound with different velocities for the longitudinal and transverse modes is assumed. As in time-of-flight diffraction, there exists, for certain scanning variables, a forbidden range in the one-phonon TDS of slower-than-sound neutrons, and this permits the determination of the sound velocity in the crystal. A fast algorithm is given for the TDS correction of neutron diffraction data collected at a fixed wavelength: this algorithm is similar to that reported earlier for the time-of-flight case. (orig.)

  10. 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.)

  11. The PTB thermal neutron reference field at GeNF

    International Nuclear Information System (INIS)

    Boettger, R.; Friedrich, H.; Janssen, H.

    2004-01-01

    The experimental setup and procedure for the characterization of the thermal neutron reference field established at the Geesthacht neutron facility (GeNF) of the GKSS is described. The neutron beam, free in air, with a maximum flux of 10 6 s -1 , can easily be reduced to less than 10 4 s -1 by using a diaphragm variable in size and without changing the beam divergence. Also, the spectral distribution with a mean energy of 45 meV, measured by time-of-flight over a 6.6 m long flight path, is independent of the beam current chosen. In the 2002/2003 experiments reported here, a 6 Li glass detector was employed to determine the absolute beam current and to calibrate the 3 He transmission beam monitor. In addition, activation measurements of gold foils were carried out at a reduced beam divergence. The results agree within ±2%. Furthermore, the beam is characterized by a low gamma background intensity and a negligible fraction of epithermal neutrons. Irradiations in combination with a scanner device to achieve a homogeneously illuminated scan field have shown that the thermal beam is well suited for dosemeter development and for the calibration of radiation protection instruments. (orig.)

  12. The PTB thermal neutron reference field at GeNF

    Energy Technology Data Exchange (ETDEWEB)

    Boettger, R.; Friedrich, H.; Janssen, H.

    2004-07-01

    The experimental setup and procedure for the characterization of the thermal neutron reference field established at the Geesthacht neutron facility (GeNF) of the GKSS is described. The neutron beam, free in air, with a maximum flux of 10{sup 6} s{sup -1}, can easily be reduced to less than 10{sup 4} s{sup -1} by using a diaphragm variable in size and without changing the beam divergence. Also, the spectral distribution with a mean energy of 45 meV, measured by time-of-flight over a 6.6 m long flight path, is independent of the beam current chosen. In the 2002/2003 experiments reported here, a {sup 6}Li glass detector was employed to determine the absolute beam current and to calibrate the {sup 3}He transmission beam monitor. In addition, activation measurements of gold foils were carried out at a reduced beam divergence. The results agree within {+-}2%. Furthermore, the beam is characterized by a low gamma background intensity and a negligible fraction of epithermal neutrons. Irradiations in combination with a scanner device to achieve a homogeneously illuminated scan field have shown that the thermal beam is well suited for dosemeter development and for the calibration of radiation protection instruments. (orig.)

  13. Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

    International Nuclear Information System (INIS)

    Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

    2015-01-01

    The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

  14. Study of the temperature evolution of defect agglomerates in neutron irradiated molybdenum single crystals

    International Nuclear Information System (INIS)

    Lambri, O.A.; Zelada-Lambri, G.I.; Cuello, G.J.; Bozzano, P.B.; Garcia, J.A.

    2009-01-01

    Small angle neutron scattering as a function of temperature, differential thermal analysis, electrical resistivity and transmission electron microscopy studies have been performed in low rate neutron irradiated single crystalline molybdenum, at room temperature, for checking the evolution of the defects agglomerates in the temperature interval between room temperature and 1200 K. The onset of vacancies mobility was found to happen in temperatures within the stage III of recovery. At around 550 K, the agglomerates of vacancies achieve the largest size, as determined from the Guinier approximation for spherical particles. In addition, the decrease of the vacancy concentration together with the dissolution of the agglomerates at temperatures higher than around 920 K was observed, which produce the release of internal stresses in the structure.

  15. Study of the temperature evolution of defect agglomerates in neutron irradiated molybdenum single crystals

    Energy Technology Data Exchange (ETDEWEB)

    Lambri, O.A. [Instituto de Fisica Rosario. Member of the CONICET' s Research Staff, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina); Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina)], E-mail: olambri@fceia.unr.edu.ar; Zelada-Lambri, G.I. [Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina); Cuello, G.J. [Institut Laue Langevin, 6, rue Jules Horowitz, BP 156, 38042 Grenoble (France); Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain); Bozzano, P.B. [Laboratorio de Microscopia Electronica. Unidad de Actividad Materiales, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Avda. Gral. Paz 1499, (1650) San Martin (Argentina); Garcia, J.A. [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain)

    2009-04-15

    Small angle neutron scattering as a function of temperature, differential thermal analysis, electrical resistivity and transmission electron microscopy studies have been performed in low rate neutron irradiated single crystalline molybdenum, at room temperature, for checking the evolution of the defects agglomerates in the temperature interval between room temperature and 1200 K. The onset of vacancies mobility was found to happen in temperatures within the stage III of recovery. At around 550 K, the agglomerates of vacancies achieve the largest size, as determined from the Guinier approximation for spherical particles. In addition, the decrease of the vacancy concentration together with the dissolution of the agglomerates at temperatures higher than around 920 K was observed, which produce the release of internal stresses in the structure.

  16. Shape Isomer in 236U Populated by Thermal Neutron Capture

    DEFF Research Database (Denmark)

    Andersen, Verner; Christensen, Carl Jørgen; Borggreen, J.

    1976-01-01

    The 116 ns shape isomer in 236U was populated by thermal neutron capture. Conversion electrons and X-rays were detected simultaneously in delayed coincidence with fission. The ratio of delayed to prompt fission was measured with the result, σIIf/σf = (1.0±0.2) × 10−5. A branching of the isomeric ...... decay σIIγ/σIIf = 7±2 was deduced from this number. No definite electron line structure was observed....

  17. An analytical method for neutron thermalization calculations in heterogenous reactors

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1965-07-01

    It is well known that the use of the diffusion approximation for stuneutron thermalization in . heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations.

  18. Activation experiment for concrete blocks using thermal neutrons

    Science.gov (United States)

    Okuno, Koichi; Tanaka, Seiichiro

    2017-09-01

    Activation experiments for ordinary concrete, colemanite-peridotite concrete, B4C-loaded concrete, and limestone concrete are carried out using thermal neutrons. The results reveal that the effective dose for gamma rays from activated nuclides of colemanite-peridotite concrete is lower than that for the other types of concrete. Therefore, colemanite-peridotite concrete is useful for reducing radiation exposure for workers.

  19. An analytical method for neutron thermalization calculations in heterogenous reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1965-01-01

    It is well known that the use of the diffusion approximation for studying neutron thermalization in heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations

  20. Attenuation of thermal neutrons by an imperfect single crystal

    Energy Technology Data Exchange (ETDEWEB)

    Naguib, K.; Adib, M. [National Research Centre, Cairo (Egypt). Reactor and Neutron Physics Dept.

    1996-06-14

    A semi-empirical formula is given which allows one to calculate the total thermal cross section of an imperfect single crystal as a function of crystal constants, temperature and neutron energy E, in the energy range between 3 meV and 10 eV. The formula also includes the contribution of the parasitic Bragg scattering to the total cross section that takes into account the crystal mosaic spread value and its orientation with respect to the neutron beam direction. A computer program (ISCANF) was developed to calculate the total attenuation of neutrons using the proposed formula. The ISCANF program was applied to investigate the neutron attenuation through a copper single crystal. The calculated values of the neutron transmission through the imperfect copper single crystal were fitted to the measured ones in the energy range 3-40 meV at different crystal orientations. The result of fitting shows that use of the computer program ISCANF allows one to predict the behaviour of the total cross section of an imperfect copper single crystal for the whole energy range. (author).

  1. Attenuation of thermal neutrons by an imperfect single crystal

    Science.gov (United States)

    Naguib, K.; Adib, M.

    1996-06-01

    A semi-empirical formula is given which allows one to calculate the total thermal cross section of an imperfect single crystal as a function of crystal constants, temperature and neutron energy E, in the energy range between 3 meV and 10 eV. The formula also includes the contribution of the parasitic Bragg scattering to the total cross section that takes into account the crystal mosaic spread value and its orientation with respect to the neutron beam direction. A computer program (ISCANF) was developed to calculate the total attenuation of neutrons using the proposed formula. The ISCANF program was applied to investigate the neutron attenuation through a copper single crystal. The calculated values of the neutron transmission through the imperfect copper single crystal were fitted to the measured ones in the energy range 3 - 40 meV at different crystal orientations. The result of fitting shows that use of the computer program ISCANF allows one to predict the behaviour of the total cross section of an imperfect copper single crystal for the whole energy range.

  2. distributions for the thermal neutron induced fission of 234U

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2016-01-01

    In addition, the analysis of thermal neutron induced fission of 234U(n,f will be discussed. Currently analysis of data is ongoing, originally taken at the ILL reactor. The experiment is of particular interest since no measurement exist of the mass and energy distributions for this system at thermal energies. One main problem encountered during analysis was the huge background of 235U(nth,f. Despite the negligible isotopic traces in the sample, the cross section difference is enormous. Solution to this parasitic background will be highlighted.

  3. LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  5. The determination of self-powered neutron detector sensitivity on thermal and epithermal neutron flux densities

    International Nuclear Information System (INIS)

    Erben, O.

    1980-01-01

    The coefficients of thermal and epithermal neutron flux density depression and self-shielding for the SPN detectors with vanadium, rhodium, silver and cobalt emitters are presented, (for cobalt SPN detectors the functions describing the absorbtion of neutrons along the emitter cross-section are also shown). Using these coefficients and previously published beta particle escape efficiencies, sensitivities are determined for the principal types of detectors produced by Les Cables de Lyon and SODERN companies. The experiments and their results verifying the validity of the theoretical work are described. (author)

  6. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs

  7. Transparent lithiated polymer films for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Mabe, Andrew N., E-mail: andrew.n.mabe@gmail.com [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Auxier, John D. [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Urffer, Matthew J. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Penumadu, Dayakar [Department of Civil and Environmental Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Schweitzer, George K. [Department of Chemistry, University of Tennessee, Knoxville, TN 37996 (United States); Miller, Laurence F. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2013-09-11

    Novel water-soluble {sup 6}Li loaded copolymer scintillation films have been designed and fabricated to detect thermal neutrons. Styrene and maleic anhydride were copolymerized to form an alternating copolymer, then the anhydride functionality was hydrolyzed using {sup 6}Li hydroxide. The resulting poly(styrene-co-lithium maleate) was mixed with salicylic acid as a fluor and cast as a thin film from water. The maximum {sup 6}Li loading obtained that resulted in a transparent film was 4.36% by mass ({sup 6}Li to polymer). The optimum fluorescence output was obtained for 11.7% salicylic acid by mass, presumably in the form of lithium salicylate, resulting in an optimum film containing 3.85% by mass of {sup 6}Li. A facile and robust synthesis method, film fabrication protocol, photoluminescence results, and scintillation responses are reported herein. -- Highlights: • A transparent polymer scintillator containing 3.85 wt% {sup 6}Li has been synthesized. • This class of polymeric thermal neutron scintillation detector is water-soluble. • Salicylic acid, presumably in the form of lithium salicylate, is used as a fluor. • The material emits 373 photons/α ({sup 241}Am) and an average of 139 photons/β ({sup 36}Cl). • The material emits 360 photons per thermal neutron capture event.

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

    International Nuclear Information System (INIS)

    Ritenour, R.L.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. Integral Parameters of the Thermal Neutron Scattering Law

    International Nuclear Information System (INIS)

    Purohit, S.N.

    1964-09-01

    Integral parameters of the thermal neutron scattering law - the thermalization binding parameter (M 2 ), the Placzek's moments of the generalized frequency spectrum of dynamical modes and the energy transfer moments of the scattering law - are theoretically discussed. A detailed study of the variation of M 2 , the thermalization time constant and the effective temperature of the vibrating atoms, with the relative weight between intra-molecular vibrations and hindered rotations for H 2 O, is presented. Theoretical results for different scattering models of H 2 O are compared with the measurements of integral experiments. A set of integral parameters for D 2 O, using Butler's model, have been obtained. Importance of the structure of hindered rotations of H 2 O and D 2 O in the study of integral parameters has also been discussed

  11. Integral Parameters of the Thermal Neutron Scattering Law

    Energy Technology Data Exchange (ETDEWEB)

    Purohit, S N

    1964-09-15

    Integral parameters of the thermal neutron scattering law - the thermalization binding parameter (M{sub 2}), the Placzek's moments of the generalized frequency spectrum of dynamical modes and the energy transfer moments of the scattering law - are theoretically discussed. A detailed study of the variation of M{sub 2}, the thermalization time constant and the effective temperature of the vibrating atoms, with the relative weight between intra-molecular vibrations and hindered rotations for H{sub 2}O, is presented. Theoretical results for different scattering models of H{sub 2}O are compared with the measurements of integral experiments. A set of integral parameters for D{sub 2}O, using Butler's model, have been obtained. Importance of the structure of hindered rotations of H{sub 2}O and D{sub 2}O in the study of integral parameters has also been discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-11

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. Microstructural evolution of martensitic steels during fast neutron iradiation

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1989-01-01

    Irradiation of martensitic/ferritic steels with fast neutrons (E > 0.1 MeV) to displacement damage levels of 30--50 dpa at temperatures of 300--500 degree C produces significant changes in the as-tempered microstructure. Dislocation loops and networks can be produced, irradiation-induced precipitates can form, the lath/subgrain boundary structure and the thermal precipitates produced during tempering can become unstable, and if helium is present, bubbles and voids can form. These microstructural changes caused by irradiation can have important effects on the properties of this class of steels for both fast breeder reactor (FBR) and magnetic fusion reactor (MFR) applications. The purpose of this paper is to compare reactor-irradiated and long-term thermally aged 9Cr--1MoVNb specimens, in order to distinguish effects due to displacement damage from those caused by elevated-temperature exposure alone. 7 refs., 1 fig

  15. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  16. About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

    International Nuclear Information System (INIS)

    Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

    2004-01-01

    Full text: In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillar systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm

  17. About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

    International Nuclear Information System (INIS)

    Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

    2004-01-01

    In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillary systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm. (author)

  18. Thermal evolution of massive compact objects with dense quark cores

    International Nuclear Information System (INIS)

    Hess, Daniel; Sedrakian, Armen

    2011-01-01

    We examine the thermal evolution of a sequence of compact objects containing low-mass hadronic and high-mass quark-hadronic stars constructed from a microscopically motivated equation of state. The dependence of the cooling tracks in the temperature versus age plane is studied on the variations of the gaplessness parameter (the ratio of the pairing gap for red-green quarks to the electron chemical potential) and the magnitude of blue quark gap. The pairing in the red-green channel is modeled assuming an inhomogeneous superconducting phase to avoid tachionic instabilities and anomalies in the specific heat; the blue colored condensate is modeled as a Bardeen-Cooper-Schrieffer (BCS)-type color superconductor. We find that massive stars containing quark matter cool faster in the neutrino-cooling era if one of the colors (blue) is unpaired and/or the remaining colors (red-green) are paired in a inhomogeneous gapless superconducting state. The cooling curves show significant variations along the sequence, as the mass (or the central density) of the models is varied. This feature provides a handle for fine-tuning the models to fit the data on the surface temperatures of same-age neutron stars. In the late-time photon cooling era we observe inversion in the temperature arrangement of models, i.e., stars experiencing fast neutrino cooling are asymptotically hotter than their slowly cooling counterparts.

  19. Seed irradiation with continuously increasing doses of thermal neutrons

    International Nuclear Information System (INIS)

    Uhlik, J.; Pfeifer, M.; Pittermann, P.

    1977-01-01

    In the 'Raman' pea cv. the biological activity of thermal neutrons was investigated after irradiation of a 780 mm column of seeds for 3000 and 4167 seconds with a flux of 5.607 x 10 9 n.cm -2 per second. For different fractions of the seed column the average density of the neutron flux was calculated. It was proved that for the described method of seed irradiation it was sufficient to determine only the dose approaching the lethal dose. If a sufficiently high column of seeds is used part of the column of seeds will be irradiated with the optimum range of doses. The advantages of the suggested method of irradiation are not only smaller time and technological requirements resulting from the need for the determination of only the critical lethal dose of radiation by means of inhibition tests performed with seedlings, but also a simpler irradiation procedure. The suggested method of irradiation is at least nine times cheaper. (author)

  20. Thermal neutron group constants in monoatomic-gas approximation

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V; Bosevski, T [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-12-15

    To solve the problem of space-energy neutron distribution in an elementary reactor cell, a combination of the multigroup procedure and the P{sub 3} approximation of the spherical harmonics method was chosen. The calculation was divided into two independent parts: the first part was to provide multigroup constants which serve as input data for the second part - the determination of the slow neutron spectra. In the present report only the first part of the problem will be discussed. The velocity dependence of cross-sections and scattering function in thermal range was interpreted by the monoatomic-gas model. A digital computer program was developed for the evaluation of the group values for these quantities (author00.

  1. Measurement of thermal neutron spectra using LINAC in Japan Atomic Energy Research Institute (JAERI)

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-01-01

    The exact grasp of thermal neutron spectra in a core region is very important for obtaining accurate thermal neutron group constants in the calculation for the nuclear design of a reactor core. For the accurate grasp of thermal neutron spectra, the capability of thermal neutron spectra to describe the moderator cross-sections for thermal neutron scattering is a key factor. Accordingly, 0 deg angular thermal neutron spectra were measured by the time of flight (TOF) method using the JAERI LINAC as a pulsed neutron source, for light water system added with Cd and In, high temperature graphite system added with boron, and light water-natural uranium heterogeneous multiplication system among the reactor moderators of light water or graphite systems. First, the equations to give the time of flight and neutron flux by TOF method were analyzed, and several corrections were investigated, such as those for detector efficiency, background, the transmission coefficient of air and the Al window of a flight tube, mean emission time of neutrons, and the distortion effect of re-entrant hole on thermal neutron spectra. Then, the experimental system, results and calculation were reported for the experiments on the above three moderator systems. Finally, the measurement of fast neutron spectra in natural uranium system and that of the efficiency of a 6 Li glass scintillator detector are described. (Wakatsuki, Y.)

  2. A thermal neutron scattering law for yttrium hydride

    Science.gov (United States)

    Zerkle, Michael; Holmes, Jesse

    2017-09-01

    Yttrium hydride (YH2) is of interest as a high temperature moderator material because of its superior ability to retain hydrogen at elevated temperatures. Thermal neutron scattering laws for hydrogen bound in yttrium hydride (H-YH2) and yttrium bound in yttrium hydride (Y-YH2) prepared using the ab initio approach are presented. Density functional theory, incorporating the generalized gradient approximation (GGA) for the exchange-correlation energy, is used to simulate the face-centered cubic structure of YH2 and calculate the interatomic Hellmann-Feynman forces for a 2 × 2 × 2 supercell containing 96 atoms. Lattice dynamics calculations using PHONON are then used to determine the phonon dispersion relations and density of states. The calculated phonon density of states for H and Y in YH2 are used to prepare H-YH2 and Y-YH2 thermal scattering laws using the LEAPR module of NJOY2012. Analysis of the resulting integral and differential scattering cross sections demonstrates adequate resolution of the S(α,β) function. Comparison of experimental lattice constant, heat capacity, inelastic neutron scattering spectra and total scattering cross section measurements to calculated values are used to validate the thermal scattering laws.

  3. Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

    International Nuclear Information System (INIS)

    Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

    2013-06-01

    Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

  4. Radiography and partial tomography of wood with thermal neutrons

    Science.gov (United States)

    Osterloh, K.; Fratzscher, D.; Schwabe, A.; Schillinger, B.; Zscherpel, U.; Ewert, U.

    2011-09-01

    The effective high neutron scattering absorption coefficient of hydrogen (48.5 cm 2/g) due to the scattering allows neutrons to reveal hydrocarbon structures with more contrast than X-rays, but at the same time limits the sample size and thickness that can be investigated. Many planar shaped objects, particularly wood samples, are sufficiently thin to allow thermal neutrons to transmit through the sample in a direction perpendicular to the planar face but not in a parallel direction, due to increased thickness. Often, this is an obstacle that prevents some tomographic reconstruction algorithms from obtaining desired results because of inadequate information or presence of distracting artifacts due to missing projections. This can be true for samples such as the distribution of glue in glulam (boards of wooden layers glued together), or the course of partially visible annual rings in trees where the features of interest are parallel to the planar surface of the sample. However, it should be possible to study these features by rotating the specimen within a limited angular range. In principle, this approach has been shown previously in a study with fast neutrons [2]. A study of this kind was performed at the Antares facility of FRM II in Garching with a 2.6×10 7/cm 2 s thermal neutron beam. The limit of penetration was determined for a wooden step wedge carved from a 2 cm×4 cm block of wood in comparison to other materials such as heavy metals and Lucite as specimens rich in hydrogen. The depth of the steps was 1 cm, the height 0.5 cm. The annual ring structures were clearly detectable up to 2 cm thickness. Wooden specimens, i.e. shivers, from a sunken old ship have been subjected to tomography. Not visible from the outside, clear radial structures have been found that are typical for certain kinds of wood. This insight was impaired in a case where the specimen had been soaked with ethylene glycol. In another large sample study, a planar board made of glulam has

  5. Radiography and partial tomography of wood with thermal neutrons

    International Nuclear Information System (INIS)

    Osterloh, K.; Fratzscher, D.; Schwabe, A.; Schillinger, B.; Zscherpel, U.; Ewert, U.

    2011-01-01

    The effective high neutron scattering absorption coefficient of hydrogen (48.5 cm 2 /g) due to the scattering allows neutrons to reveal hydrocarbon structures with more contrast than X-rays, but at the same time limits the sample size and thickness that can be investigated. Many planar shaped objects, particularly wood samples, are sufficiently thin to allow thermal neutrons to transmit through the sample in a direction perpendicular to the planar face but not in a parallel direction, due to increased thickness. Often, this is an obstacle that prevents some tomographic reconstruction algorithms from obtaining desired results because of inadequate information or presence of distracting artifacts due to missing projections. This can be true for samples such as the distribution of glue in glulam (boards of wooden layers glued together), or the course of partially visible annual rings in trees where the features of interest are parallel to the planar surface of the sample. However, it should be possible to study these features by rotating the specimen within a limited angular range. In principle, this approach has been shown previously in a study with fast neutrons . A study of this kind was performed at the Antares facility of FRM II in Garching with a 2.6x10 7 /cm 2 s thermal neutron beam. The limit of penetration was determined for a wooden step wedge carved from a 2 cmx4 cm block of wood in comparison to other materials such as heavy metals and Lucite as specimens rich in hydrogen. The depth of the steps was 1 cm, the height 0.5 cm. The annual ring structures were clearly detectable up to 2 cm thickness. Wooden specimens, i.e. shivers, from a sunken old ship have been subjected to tomography. Not visible from the outside, clear radial structures have been found that are typical for certain kinds of wood. This insight was impaired in a case where the specimen had been soaked with ethylene glycol. In another large sample study, a planar board made of glulam has been

  6. Lethal Effect of Thermal Neutrons on Hypoxic Elirlich Ascites Tumour Cells in vitro

    OpenAIRE

    MITSUHIKO, AKABOSHI; KENICHI, KAWAI; HIROTOSHI, MAKI; Research Reactor Institute, Kyoto University; Research Reactor Institute, Kyoto University; Research Reactor Institute, Kyoto University

    1985-01-01

    Ehrlich ascites tumour cells were irradiated in vitro with thermal neutrons under aerobic and hypoxic conditions, and the survival of their reproductive capacity was assayed in vivo. Only a slight hypoxic protection was observed for thermal neutron irradiation with an oxygen enhancement ratio (OER) of 1.2, as compared with OER of 3.3 for ^Co-γ-rays. Absorbed dose of thermal neutrons was calculated by assuming that the energies of recoiled nuclei were completely absorbed within a cell nucleus....

  7. Carbon filter property detection with thermal neutron technique

    International Nuclear Information System (INIS)

    Deng Zhongbo; Han Jun; Li Wenjie

    2003-01-01

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

  8. Gas loop - continuous measurement of thermal and fast neutron fluxes

    International Nuclear Information System (INIS)

    Droulers, Y.; Pleyber, G.; Sciers, P.; Maurin, G.

    1964-01-01

    The measurement method described in this report can be applied both to thermal and fast neutron fluxes. A description is given of two practical applications in each of these two domains. This method is particularly suitable for measurements carried out on 'loop' type equipment. The measurement of the relative flux variations are carried out with an accuracy of 5 per cent. The choice of the shape of the gas circuit leaves a considerable amount of liberty for the adaptation of the measurement circuit to the experimental conditions. (authors) [fr

  9. Non-destructive characterization using pulsed fast-thermal neutrons

    International Nuclear Information System (INIS)

    Womble, P.C.

    1995-01-01

    Explosives, illicit drugs, and other contraband materials contain various chemical elements in quantities and ratios that differentiate them from each other and from innocuous substances. Furthermore, the major chemical elements in coal can provide information about various parameters of importance to the coal industry. In both examples, the non-destructive identification of chemical elements can be performed using pulsed fast-thermal neutrons that, through nuclear reactions, excite the nuclei of the various elements. This technique is being currently developed for the dismantling of nuclear weapons classified as trainers, and for the on-line coal bulk analysis. (orig.)

  10. An extension of diffusion theory for thermal neutrons near boundaries

    International Nuclear Information System (INIS)

    Alvarez Rivas, J. L.

    1963-01-01

    The distribution of thermal neutron flux has been measured inside and outside copper rods of several diameters, immersed in water. It has been found that these distributions can be calculated by means of elemental diffusion theory if the value of the coefficient of diffusion is changed. this parameter is truly a diffusion coefficient, which now also depends on the diameter of the rod. Through a model an expression of this coefficient is introduced which takes account of the measurements of the author and of those reported in PIGC P/928 (1995), ANL-5872 (1959), DEGR 319 (D) (1961). This model could be extended also to plane geometry. (Author) 19 refs

  11. 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

  12. Thermal design and test verification of GALAXY evolution explorer (GALEX)

    Science.gov (United States)

    Wu, P. S.; Lee, S. -C.

    2002-01-01

    This paper describes the thermal control design of GALEX, an ultraviolet telescope that investigates the UV properties of local galaxies, history of star formation, and global causes of star formation and evolution.

  13. Imaging of Rabbit VX-2 Hepatic Cancer by Cold and Thermal Neutron Radiography

    Science.gov (United States)

    Tsuchiya, Yoshinori; Matsubayashi, Masahito; Takeda, Tohoru; Lwin, Thet Thet; Wu, Jin; Yoneyama, Akio; Matsumura, Akira; Hori, Tomiei; Itai, Yuji

    2003-11-01

    Neutron radiography is based on differences in neutron mass attenuation coefficients among the elements and is a non-destructive imaging method. To investigate biomedical applications of neutron radiography, imaging of rabbit VX-2 liver cancer was performed using thermal and cold neutron radiography with a neutron imaging plate. Hepatic vessels and VX-2 tumor were clearly observed by neutron radiography, especially by cold neutron imaging. The image contrast of this modality was better than that of absorption-contrast X-ray radiography.

  14. Residual stress evolution regularity in thermal barrier coatings under thermal shock loading

    Directory of Open Access Journals (Sweden)

    Ximin Chen

    2014-01-01

    Full Text Available Residual stress evolution regularity in thermal barrier ceramic coatings (TBCs under different cycles of thermal shock loading of 1100°C was investigated by the microscopic digital image correlation (DIC and micro-Raman spectroscopy, respectively. The obtained results showed that, as the cycle number of the thermal shock loading increases, the evolution of the residual stress undergoes three distinct stages: a sharp increase, a gradual change, and a reduction. The extension stress near the TBC surface is fast transformed to compressive one through just one thermal cycle. After different thermal shock cycles with peak temperature of 1100°C, phase transformation in TBC does not happen, whereas the generation, development, evolution of the thermally grown oxide (TGO layer and micro-cracks are the main reasons causing the evolution regularity of the residual stress.

  15. 6Li-doped silicate glass for thermal neutron shielding

    International Nuclear Information System (INIS)

    Stone, C.A.; Blackburn, D.H.; Kauffman, D.A.; Cranmer, D.C.; Olmez, I.

    1994-01-01

    Glass formulations are described that contain high concentrations of 6 Li and are suitable for use as thermal neutron shielding. One formulation contained 31 mol% of 6 Li 2 O and 69 mol% of SiO 2 . Studies were performed on a second formulation that contained as much as 37 mol% of 6 Li 2 O and 59 mol% of SiO 2 , with 4 mol% Al 2 O 3 added to prevent crystallization at such high 6 Li 2 O concentrations. These lithium silicate glasses can be formed into a variety of shapes using conventional glass fabrication techniques. Examples include flat plates, disks, hollow cylinders, and other more complex geometries. Both in-beam and in-core experiments have been performed to study the use and durability of Li silicate glasses. In-core experiments show the glass can withstand the intense radiation fields near the core of a reactor. The neutron attenuation of the glasses used in these studies was 90%/mm. In-beam studies show that the glass is effective for reducing the gamma-ray and neutron fields near experiments. ((orig.))

  16. Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10)

    International Nuclear Information System (INIS)

    Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W.

    2017-01-01

    The Laboratório Brasileiro de Nêutrons do Instituto de Radioproteção e Dosimetria (IRD/CNEN) has developed and built a thermal neutron flux facility to provide neutron fluence for dosimeters (Astuto, 2014). This fluency is obtained by four 16 Ci sources 241 AmBe (α, n) positioned around the channel positioned in the center of the Thermal Flow Unit (UFT). The UFT was built with blocks of paraffin with graphite addition and graphite blocks of high purity to obtain a central field with a homogeneous thermal neutron fluence for calibration purposes with the following measurements: 1.2 x 1.2 x 1.2 m 3 . The objective of this work is to characterize several points, in the thermal energy range, in terms of the equivalent ambient dose quantity H⁎(10) for calibration and irradiation of monitors neutrons

  17. ESR-dosimetry in thermal and epithermal neutron fields for application in boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, Tobias

    2016-01-22

    Dosimetry is essential for every form of radiotherapy. In Boron Neutron Capture Therapy (BNCT) mixed neutron and gamma fields have to be considered. Dose is deposited in different neutron interactions with elements in the penetrated tissue and by gamma particles, which are always part of a neutron field. The therapeutic dose in BNCT is deposited by densely ionising particles, originating from the fragmentation of the isotope boron-10 after capture of a thermal neutron. Despite being investigated for decades, dosimetry in neutron beams or fields for BNCT remains complex, due to the variety in type and energy of the secondary particles. Today usually ionisation chambers combined with metal foils are used. The applied techniques require extensive effort and are time consuming, while the resulting uncertainties remain high. Consequently, the investigation of more effective techniques or alternative dosimeters is an important field of research. In this work the possibilities of ESR-dosimeters in those fields have been investigated. Certain materials, such as alanine, generate stable radicals upon irradiation. Using Electron Spin Resonance (ESR) spectrometry the amount of radicals, which is proportional to absorbed dose, can be quantified. Different ESR detector materials have been irradiated in the thermal neutron field of the research reactor TRIGA research reactor in Mainz, Germany, with five setups, generating different secondary particle spectra. Further irradiations have been conducted in two epithermal neutron beams. The detector response, however, strongly depends on the dose depositing particle type and energy. It is hence necessary to accompany measurements by computational modelling and simulation. In this work the Monte Carlo code FLUKA was used to calculate absorbed doses and dose components. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using amorphous track models. For the simulation, detailed models of

  18. Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

    Energy Technology Data Exchange (ETDEWEB)

    Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

    2005-11-15

    A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

  19. Supercritical accretion in the evolution of neutron star binaries and its implications

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang-Hwan, E-mail: clee@pusan.ac.kr; Cho, Hee-Suk

    2014-08-15

    Recently ∼2M{sub ⊙} neutron stars PSR J1614-2230 and PSR J0348+0432 have been observed in neutron star-white dwarf binaries. These observations ruled out many neutron star equations of states with which the maximum neutron star mass becomes less than 2M{sub ⊙}. On the other hand, all well-measured neutron star masses in double neutron star binaries are still less than 1.5M{sub ⊙}. In this article we suggest that 2M{sub ⊙} neutron stars in neutron star-white dwarf binaries are the result of the supercritical accretion onto the first-born neutron star during the evolution of the binary progenitors.

  20. Evaluation of RBE of thermal neutron capture reaction

    International Nuclear Information System (INIS)

    Fukuda, Hiroshi; Matsuzawa, Taiju; Kobayashi, Toru; Kanda, Keiji.

    1985-01-01

    B16 melanoma cells were grown in a flask (Falcon 3031). When the cells reached the latter stage of logarithmic phase, B-boric acid (92 % concentrated 10 B) was added to the flask until 5 μg/ml medium was attained (Medium I). The other medium did not contain 10 B (Medium II). After both media were exposed to thermal neutrons, survival curves were obtained from the colony method and the absorbed dose of the cells were obtained from the mathematical models. Survival curves from the colony method had no shoulders, showing that Do was 0.95 x 10 12 n/cm 2 in Medium I and 3.2 x 10 12 n/cm 2 in Medium II. Do calculated by mathematical models was 0.507 Gy in Medium I and 0.604 Gy in Medium II. REB of thermal neutrons was 3.04 in Medium I and 2.55 in Medium II. REB of 10 B (n, α) 7 Li reaction was 3.30. (Namekawa, K.)

  1. Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

    Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

  2. Study and development of new dosemeters for thermal neutrons; Estudio y desarrollo de nuevos dosimetros para neutrones termicos

    Energy Technology Data Exchange (ETDEWEB)

    Urena N, F

    1998-12-31

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation ({phi} = 5 x 10{sup 7} n/cm{sup 2} -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence {Phi} = {phi} {Delta}t ; where {Delta}t = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  3. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    2001-05-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  4. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    2001-01-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  5. Determination of the thermal and epithermal neutron sensitivities of an LBO chamber

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Satoru; Kotani, Kei; Kajimoto, Tsuyoshi; Tanaka, Kenichi [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Sato, Hitoshi; Nakajima, Erika [Ibaraki Prefectural University of Health Science, Radiological Sciences, Ibaraki (Japan); Shimazaki, Takuto [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Delta Kogyo Co., Ltd., Hiroshima (Japan); Suda, Mitsuru; Hamano, Tsuyoshi [National Institute of Radiological Sciences, Chiba-Shi, Chiba (Japan); Hoshi, Masaharu [Hiroshima University, Institute for Peace Science, Hiroshima (Japan)

    2017-08-15

    An LBO (Li{sub 2}B{sub 4}O{sub 7}) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the {sup 9}Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm{sup 2}, i.e., from the chamber response divided by neutron fluence (cm{sup -2}). The measured LBO chamber sensitivities were 2.23 x 10{sup -7} ± 0.34 x 10{sup -7} (pC cm{sup 2}) for thermal neutrons and 2.00 x 10{sup -5} ± 0.12 x 10{sup -5} (pC cm{sup 2}) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation. (orig.)

  6. Miniature neutron sources: Thermal neutron sources and their users in the academic field

    International Nuclear Information System (INIS)

    Egelstaff, P.A.

    1992-01-01

    The three levels of thermal neutron sources are introduced - University laboratory sources infrastructure sources and world-class sources - and the needs for each kind and their inter-dependence will be emphasized. A description of the possibilities for University sources based on α-Be reactions or spontaneous fission emission is given, and current experience with them is described. A new generation of infrastructure sources is needed to continue the regional programs based on small reactors. Some possibilities for accelerator sources that could meet this need are considered

  7. Calculation of the thermal neutron flux depression in the loop VISA-1

    International Nuclear Information System (INIS)

    Martinc, R.

    1961-01-01

    Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr

  8. Histological and Physiological Alterations Induced by Thermal Neutron Fluxes in Male Swiss Albino Mice

    International Nuclear Information System (INIS)

    Alzergy, A.A.; Emara, N.M.; Abd El-Latif, A.A.; El-Saady, S.M.M.; Emara, N.M.; Abd El-Latif, A.A.

    2010-01-01

    This work was performed to investigate the biological effects of different thermal neutron fluxes (0.27x10 8 , 0.52X10 8 , 1.089X10 8 , 2.16X10 8 and 4.32X10 8 ) on liver and kidney of male mice using neutron irradiation cell with Ra-Be(α,n) 3 mCi neutron source Leybold (55930). Exposed to various fluxes of thermal neutron induced a dramatic alterations in hepatic and renal functions as indicated by biochemical estimation of several parameters (bilirubin, SGT, and alkaline phosphate .Urea , total protein, and albumin) and confirmed by histological examinations Thermal neutron exposure induces marked increase in the serum activities of total bilirubin, alanine amino transaminase (ALT or GPT), and alkaline phosphate, whereas, urea, total protein and albumin showed marked decline as compared to control group. The physiological changes induced in thermal neutron fluxes dependent manner. Histopathological results revealed mild to severe type of necrosis, and degenerative changes in liver and kidney of male mice exposed to thermal neutron fluxes. Also it was found that the histopathological alterations induced in thermal neutron fluxes dependent manner. It was found that exposed to thermal neutron fluxes irradiation plays prominent role in the development of the physiological alterations in male Swiss albino mice. The Former up normalities as a result of the sequence events followed interaction of radiation with the former biological mater (liver and kidney) of male Swiss albino mice, which are, physical, physicochemical, chemical, and biological stages.

  9. Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

    2016-01-01

    In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

  10. Filtered thermal neutron captured cross sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Pham Ngoc Son; Vuong Huu Tan

    2015-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

  11. Cross-section of single-crystal materials used as thermal neutron filters

    International Nuclear Information System (INIS)

    Adib, M.

    2005-01-01

    Transmission properties of several single crystal materials important for neutron scattering instrumentation are presented. A computer codes are developed which permit the calculation of thermal diffuse and Bragg-scattering cross-sections of silicon., and sapphire as a function of material's constants, temperature and neutron energy, E, in the range 0.1 MeV .A discussion of the use of their single-crystal as a thermal neutron filter in terms of the optimum crystal thickness, mosaic spread, temperature, cutting plane and tuning for efficient transmission of thermal-reactor neutrons is given

  12. Thermal neutron detection by means of an organic solid-state track detector

    International Nuclear Information System (INIS)

    Doerschel, B.; Streubel, G.

    1979-01-01

    Thermal neutrons can be detected by means of organic solid-state track detectors if they are combined with radiators in which charged secondary particles are produced in neutron interaction processes. The secondary particles can produce etchable tracks in the detector material. For thermal neutron fluence determination from the track densities, the thermal neutron sensitivity was calculated for cellulose triacetate detectors with LiF radiators, taking into account energy and angular distribution of the alpha particles produced in the LiF radiator. This value is in good agreement with the sensitivity measured during irradiation in different neutron fields if corrections are considered the production of etchable or visuable tracks. Measuring range and measuring accuracy meet the requirements of thermal neutron detection in personnel dosimetry. Possibilities of extending the measuring range are discussed. (author)

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

    International Nuclear Information System (INIS)

    Shaaban, I.

    2009-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  15. Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

    International Nuclear Information System (INIS)

    Andrejsek, K.

    1977-01-01

    The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

  16. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  17. Investigation of coupling scheme for neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

    1988-01-01

    Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

  18. A new position-sensitive detector for thermal and epithermal neutrons

    International Nuclear Information System (INIS)

    Jeavons, A.P.; Ford, N.L.; Lindberg, B.; Sachot, R.

    1977-01-01

    A new two-dimensional position-sensitive neutron detector is described. It is based on (n,γ) neutron resonance capture in a foil with subsequent detection of internal conversion electrons with a high-density proportional chamber. Large-area detectors with a 1 mm spatial resolution are feasible. A detection efficiency of 50% is possible for thermal neutrons using gadolinium-157 foil and for epithermal neutrons using hafnium-177. (Auth.)

  19. Analysis of the Photoneutron Yield and Thermal Neutron Flux in an Unreflected Electron Accelerator-Driven Neutron Source

    International Nuclear Information System (INIS)

    Dale, Gregory E.; Gahl, John M.

    2005-01-01

    There are several potential uses for a high-flux thermal neutron source in both industrial and clinical applications. The viable commercial implementation of these applications requires a low-cost, high-flux thermal neutron generator suitable for installation in industrial and clinical environments. This paper describes the Monte Carlo for N-Particle modeling results of a high-flux thermal neutron source driven with an electron accelerator. An electron linear accelerator (linac), fitted with a standard X-ray converter, can produce high neutron yields in materials with low photonuclear threshold energies, such as D and 9 Be. Results indicate that a 10-MeV, 10-kW electron linac can produce on the order of 10 12 n/s in a heavy water photoneutron target. The thermal neutron flux in an unreflected heavy water target is calculated to be on the order of 10 10 n.cm -2 .s. The sensitivity of these answers to heavy water purity is also investigated, specifically the dilution of heavy water with light water. It is shown that the peak thermal neutron flux is not adversely effected by dilution up to a light water weight fraction of 35%

  20. Flux distribution in phantom for biomedical use of beam-type thermal neutrons

    International Nuclear Information System (INIS)

    Aoki, Kazuhiko; Kobayashi, Tooru; Kanda, Keiji; Kimura, Itsuro

    1985-01-01

    For boron neutron capture therapy, the thermal neutron beam is worth using as therapeutic neutron irradiation without useless and unfavorable exposure of normal tissues around tumor and for microanalysis system to measure ppm-order 10 B concentrations in tissue and to search for the location of the metastasis of tumor. In the present study, the thermal neutron flux distribution in a phantom, when beam-type thermal neutrons were incident on it, was measured at the KUR Neutron Guide Tube. The measurements were carried out by two different methods using indium foil. The one is an ordinary foil activation technique by using the 115 In(n, γ) 116m 1 In reactions, while the other is to detect γ-rays from the 115 In(n, γ) 116m 2 In reactions during neutron irradiations with a handy-type Ge detector. The calculations with DOT 3.5 were performed to examine thermal neutron flux in the phantom for various beam size and phantom size. The experimental and calculated results are in good agreement and it is shown that the second type measurement has a potential for practical application as a new monitoring system of the thermal neutron flux in a living body for boron neutron capture therapy. (author)

  1. Visualization and measurement by image processing of thermal hydraulic phenomena by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, Nobuyuki

    1996-01-01

    Neutron Radiography was applied to visualization of thermal hydraulic phenomena and measurement was carried out by image processing the visualized images. Since attenuation of thermal neutron rays is high in ordinary liquids like water and organic fluid while it is low in most of metals, liquid flow behaviors can be visualized through a metallic wall by neutron radiography. Measurement of void fraction and flow vector field which is important to study thermal hydraulic phenomena can be carried out by image processing the images obtained by the visualization. Various two-phase and liquid metal flows were visualized by a JRR-3M thermal neutron radiography system in the present study. Multi-dimensional void fraction distributions in two-phase flows and flow vector fields in liquid metals, which are difficult to measure by the other methods, were successfully measured by image processing. It was shown that neutron radiography was efficiently applicable to study thermal hydraulic phenomena. (author)

  2. Neutron beams. Tracks analysis, imaging and medicine

    International Nuclear Information System (INIS)

    Pepy, G.

    2006-01-01

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

  3. A time-of-flight detector for thermal neutrons from radiotherapy Linacs

    Energy Technology Data Exchange (ETDEWEB)

    Conti, V. [Universita degli Studi di Milano and INFN di Milano (Italy)], E-mail: conti.Valentina@gmail.com; Bartesaghi, G. [Universita degli Studi di Milano and INFN di Milano (Italy); Bolognini, D.; Mascagna, V.; Perboni, C.; Prest, M.; Scazzi, S. [Universita dell' Insubria, Como and INFN di Milano (Italy); Mozzanica, A. [Universita degli Studi di Brescia and INFN sezione di Pavia (Italy); Cappelletti, P.; Frigerio, M.; Gelosa, S.; Monti, A.; Ostinelli, A. [Fisica Sanitaria, Ospedale S. Anna di Como (Italy); Giannini, G.; Vallazza, E. [INFN, sezione di Trieste and Universita degli Studi di Trieste (Italy)

    2007-10-21

    Boron Neutron Capture Therapy (BNCT) is a therapeutic technique exploiting the release of dose inside the tumour cell after a fission of a {sup 10}B nucleus following the capture of a thermal neutron. BNCT could be the treatment for extended tumors (liver, stomach, lung), radio-resistant ones (melanoma) or tumours surrounded by vital organs (brain). The application of BNCT requires a high thermal neutron flux (>5x10{sup 8}ncm{sup -2}s{sup -1}) with the correct energy spectrum (neutron energy <10keV), two requirements that for the moment are fulfilled only by nuclear reactors. The INFN PhoNeS (Photo Neutron Source) project is trying to produce such a neutron beam with standard radiotherapy Linacs, maximizing with a dedicated photo-neutron converter the neutrons produced by Giant Dipole Resonance by a high energy (>8MeV) photon beam. In this framework, we have developed a real-time detector to measure the thermal neutron time-of -flight to compute the flux and the energy spectrum. Given the pulsed nature of Linac beams, the detector is a single neutron counting system made of a scintillator detecting the photon emitted after the neutron capture by the hydrogen nuclei. The scintillator signal is sampled by a dedicated FPGA clock thus obtaining the exact arrival time of the neutron itself. The paper will present the detector and its electronics, the feasibility measurements with a Varian Clinac 1800/2100CD and comparison with a Monte Carlo simulation.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. 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.

  6. Thermal-neutron multiple scattering: critical double scattering

    International Nuclear Information System (INIS)

    Holm, W.A.

    1976-01-01

    A quantum mechanical formulation for multiple scattering of thermal-neutrons from macroscopic targets is presented and applied to single and double scattering. Critical nuclear scattering from liquids and critical magnetic scattering from ferromagnets are treated in detail in the quasielastic approximation for target systems slightly above their critical points. Numerical estimates are made of the double scattering contribution to the critical magnetic cross section using relevant parameters from actual experiments performed on various ferromagnets. The effect is to alter the usual Lorentzian line shape dependence on neutron wave vector transfer. Comparison with corresponding deviations in line shape resulting from the use of Fisher's modified form of the Ornstein-Zernike spin correlations within the framework of single scattering theory leads to values for the critical exponent eta of the modified correlations which reproduce the effect of double scattering. In addition, it is shown that by restricting the range of applicability of the multiple scattering theory from the outset to critical scattering, Glauber's high energy approximation can be used to provide a much simpler and more powerful description of multiple scattering effects. When sufficiently close to the critical point, it provides a closed form expression for the differential cross section which includes all orders of scattering and has the same form as the single scattering cross section with a modified exponent for the wave vector transfer

  7. Effective thermal neutron absorption cross section for heterogeneous mixture

    International Nuclear Information System (INIS)

    Gabanska, B.; Igielski, A.; Krynicka-Drozdowicz, E.; Woznicka, U.

    1989-01-01

    The first estimations (basing on Umiastowski's theory) of the influence of the sample heterogeneity of the effective thermal neutron absorption cross section were compared with the results obtained for the homogeneous mixture which components and concentration were the same as those of the heterogeneous sample. An experiment was prepared to determine how good this estimate is. Three artificial heterogeneous cylindrical samples (2R = H = 9 cm) were manufactured from pure silver cylinders embedded in plexiglass, keeping the Ag content and varying the size of cylinders (2R = H = 1.0 cm, 0.6 cm and 0.4 cm). Calculations performed show that the experimental effect of the sample heterogeneity can be significant. 5 figs., 5 tabs, 11 refs. (author)

  8. Stability of thermally induced copper precipitates under neutron irradiation

    International Nuclear Information System (INIS)

    Phythian, W.J.; Dumbill, S.; Brown, P.; Sinclair, R.

    1993-01-01

    Model Fe 1.3%Cu and Fe 1.3%Cu 1.1%Ni alloys have been thermally aged at 550 C for 2 hours (peak) and 10 hours prior to irradiation at 288 C to a dose of 5.10 22 n/m 2 . Results of a microstructural investigation using dedicated field emission gun scanning transmission electron microscopy (FEGSTEM) and small angle neutron scattering (SANS) to assess precipitate stability in the binary alloy, are presented. These data are then used to predict a hardness change as a result of copper precipitation for comparison with the measured values obtained using standard 5 kg Vickers hardness tests on the SANS samples. Implications of these data to the re-embrittlement of the RPV by subsequent copper precipitation is discussed. (authors). 16 refs., 5 figs., 5 tabs

  9. Stability of thermally induced copper precipitates under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Phythian, W J; Dumbill, S; Brown, P; Sinclair, R [AEA Technology, Harwell (United Kingdom)

    1994-12-31

    Model Fe 1.3%Cu and Fe 1.3%Cu 1.1%Ni alloys have been thermally aged at 550 C for 2 hours (peak) and 10 hours prior to irradiation at 288 C to a dose of 5.10{sup 22} n/m{sup 2}. Results of a microstructural investigation using dedicated field emission gun scanning transmission electron microscopy (FEGSTEM) and small angle neutron scattering (SANS) to assess precipitate stability in the binary alloy, are presented. These data are then used to predict a hardness change as a result of copper precipitation for comparison with the measured values obtained using standard 5 kg Vickers hardness tests on the SANS samples. Implications of these data to the re-embrittlement of the RPV by subsequent copper precipitation is discussed. (authors). 16 refs., 5 figs., 5 tabs.

  10. Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

    International Nuclear Information System (INIS)

    Filliatre, P.; Oriol, L.; Jammes, C.; Vermeeren, L.

    2009-01-01

    A fission chamber with a 242 Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux (∼10 14 ncm -2 s -1 or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a 242 Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

  11. Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

    Energy Technology Data Exchange (ETDEWEB)

    Filliatre, P. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)], E-mail: philippe.filliatre@cea.fr; Oriol, L.; Jammes, C. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)

    2009-05-21

    A fission chamber with a {sup 242}Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux ({approx}10{sup 14}ncm{sup -2}s{sup -1} or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a {sup 242}Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

  12. The influence of thermal pressure on equilibrium models of hypermassive neutron star merger remnants

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, J. D.; Ott, C. D.; Roberts, L. [TAPIR, California Institute of Technology, Mailcode 350-17, Pasadena, CA 91125 (United States); O' Connor, E. P. [CITA, University of Toronto, 60 St. George Street, Toronto, ON M5S 3H8 (Canada); Kiuchi, K. [Yukawa Institute for Theoretical Physics, University of Kyoto, Kyoto (Japan); Duez, M., E-mail: cott@tapir.caltech.edu [Department of Physics and Astronomy, Washington State University, Pullman, WA (United States)

    2014-07-20

    The merger of two neutron stars leaves behind a rapidly spinning hypermassive object whose survival is believed to depend on the maximum mass supported by the nuclear equation of state (EOS), angular momentum redistribution by (magneto-)rotational instabilities, and spindown by gravitational waves. The high temperatures (∼5-40 MeV) prevailing in the merger remnant may provide thermal pressure support that could increase its maximum mass and, thus, its life on a neutrino-cooling timescale. We investigate the role of thermal pressure support in hypermassive merger remnants by computing sequences of spherically symmetric and axisymmetric uniformly and differentially rotating equilibrium solutions to the general-relativistic stellar structure equations. Using a set of finite-temperature nuclear EOS, we find that hot maximum-mass critically spinning configurations generally do not support larger baryonic masses than their cold counterparts. However, subcritically spinning configurations with mean density of less than a few times nuclear saturation density yield a significantly thermally enhanced mass. Even without decreasing the maximum mass, cooling and other forms of energy loss can drive the remnant to an unstable state. We infer secular instability by identifying approximate energy turning points in equilibrium sequences of constant baryonic mass parameterized by maximum density. Energy loss carries the remnant along the direction of decreasing gravitational mass and higher density until instability triggers collapse. Since configurations with more thermal pressure support are less compact and thus begin their evolution at a lower maximum density, they remain stable for longer periods after merger.

  13. The influence of thermal pressure on equilibrium models of hypermassive neutron star merger remnants

    International Nuclear Information System (INIS)

    Kaplan, J. D.; Ott, C. D.; Roberts, L.; O'Connor, E. P.; Kiuchi, K.; Duez, M.

    2014-01-01

    The merger of two neutron stars leaves behind a rapidly spinning hypermassive object whose survival is believed to depend on the maximum mass supported by the nuclear equation of state (EOS), angular momentum redistribution by (magneto-)rotational instabilities, and spindown by gravitational waves. The high temperatures (∼5-40 MeV) prevailing in the merger remnant may provide thermal pressure support that could increase its maximum mass and, thus, its life on a neutrino-cooling timescale. We investigate the role of thermal pressure support in hypermassive merger remnants by computing sequences of spherically symmetric and axisymmetric uniformly and differentially rotating equilibrium solutions to the general-relativistic stellar structure equations. Using a set of finite-temperature nuclear EOS, we find that hot maximum-mass critically spinning configurations generally do not support larger baryonic masses than their cold counterparts. However, subcritically spinning configurations with mean density of less than a few times nuclear saturation density yield a significantly thermally enhanced mass. Even without decreasing the maximum mass, cooling and other forms of energy loss can drive the remnant to an unstable state. We infer secular instability by identifying approximate energy turning points in equilibrium sequences of constant baryonic mass parameterized by maximum density. Energy loss carries the remnant along the direction of decreasing gravitational mass and higher density until instability triggers collapse. Since configurations with more thermal pressure support are less compact and thus begin their evolution at a lower maximum density, they remain stable for longer periods after merger.

  14. THERMAL: A routine designed to calculate neutron thermal scattering. Revision 1

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1995-01-01

    THERMAL is designed to calculate neutron thermal scattering that is elastic and isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the relative system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy, e.g., the keV energy range. The THERMAL method is simple, clean, easy to understand, and most important very efficient; on a SUN SPARC-10 workstation, at low energies with thermal scattering it can do almost 6 million scatters a minute and at high energy over 13 million. Warning: This version of THERMAL completely supersedes the original version described in the same report number, dated February 24, 1995. The method used in the original code is incorrect, as explained in this report

  15. Neutron-induced damage evolution under Beam Raster Scanner conditions for IFMIF

    International Nuclear Information System (INIS)

    Mota, Fernando; Ortiz, Christophe J.; Ibarra, Angel; Vila, Rafael

    2011-01-01

    The formation and evolution of defects in materials irradiated with a homogeneous neutron source and with the Beam Raster Scanner (BRS) solution was investigated. The intensity neutron source fluctuations inherent to the BRS system were determined using the neutron transport McDeLicious code. Defects generated during irradiation were calculated using the binary collision approximation MARLOWE code, using the primary knock-on atom (PKA) energy spectrum resulting from neutron interactions with the material. In order to predict the evolution of defects during irradiation, a Rate Theory model based on ab initio parameters was developed. Our model accounts for the migration of mobile defects, the formation of clusters and their recombination. As an example, we investigated defect evolution in Fe irradiated at room temperature in both beam configurations. Simulation results clearly indicate that the defect evolution expected in the BRS configuration is nearly the same as the one expected in a homogeneous irradiation system.

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

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  17. Characterization of the internal background for thermal and fast neutron detection with CLLB

    Energy Technology Data Exchange (ETDEWEB)

    Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F.; Wulf, Eric A.

    2016-12-01

    We report on a set of experiments conducted to determine what effects, if any, the internal background in the CLLB scintillation detector has on the thermal neutron detection performance. We conducted source measurements using an unmoderated and moderated {sup 252}Cf neutron/γ-ray source and long (48-h), unshielded and shielded, background measurements to characterize the internal background with and without a source present. These measurements allowed us to determine the 2-d event selections needed to isolate the thermal neutron peak observed in pulse shape vs. energy space and apply those selections to our background measurements. Our results indicate that the thermal neutron detection capabilities of the CLLB are marginally affected by the presence of internal background. An unmoderated 113-µCi {sup 252}Cf source at 15 cm from the detector yields a thermal neutron rate of 8×10{sup −2}/s cm{sup 3}, while moderating the source with 5 cm of polyethylene yields a thermal neutron rate of 5.5×10{sup −1}/s cm{sup 3}. The measured background rate for events that fall within the selected thermal neutron region is 1.2×10{sup −3}/s cm{sup 3}. Lastly, the potential for CLLB for detecting fast neutrons was investigated.

  18. 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

  19. Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, D. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); INAIL – DIT, Via di Fontana Candida n.1, 00040 Monteporzio Catone (Italy); Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Bortot, D. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Palomba, M. [ENEA Casaccia, Via Anguillarese, 301, S. Maria di Galeria, 00123 Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); Gentile, A. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Strigari, L. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Pressello, C. [Department of Medical Physics, Azienda Ospedaliera San Camillo Forlanini, Circonvallazione Gianicolense 87, 00152 Roma (Italy); Soriani, A. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Gómez-Ros, J.M. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2015-10-21

    A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 10{sup 9}–10{sup 10} cm{sup −2}, using the ex-core radial thermal neutron column of the ENEA Casaccia – TRIGA reactor. The results are presented in this work.

  20. Thermal neutron cross section measurements for technetium-99

    International Nuclear Information System (INIS)

    Yates, M.A.; Schroeder, N.C.; Fowler, M.M.

    1993-01-01

    Technetium, because of its long half-like (213,000 years) and ability to migrate in the environment, is a primary contributor to the long-term radioactivity related risk associated with geologic nuclear waste disposal. One proposal for converting technetium to an environmentally benign element investigating transmutation with an accelerator-based system, (i.e., Accelerator Transmutation of Waste, ATW). Planning for efficient processing of technetium through the transmuter will require knowledge of the thermal neutron cross section for the 99 Tc (n,γ) 100 Tc reaction. The authors have recently remeasured this cross section. Weighed aliquots (19-205 μg) of a NIST traceable 99 Tc standard were irradiated for 30-150 sec using the pneumatic open-quotes rabbitclose quotes system of LANL's Omega West Reactor. The two gamma rays from the 15.7-sec half-life product were measured immediately after irradiation on a high-resolution Ge detector. Thermal fluxes were measured using gold foils and Cd wrapped gold foils. The observation cross section is 19 ± 1 b. This agrees well with the 1977 value but has half the uncertainty

  1. The chemical consequences of thermal neutron capture in alkali selenates

    International Nuclear Information System (INIS)

    Duplatre, G.; Vargas, J.I.

    1977-01-01

    The initial retention of the SeO 4 2- ion after thermal neutron capture has been studied in various matrices by chemical analysis. A comparison between the thermal behaviour of the chemically analyzed Sesup(IV) and the disappearance of the E.P.R. species SeO 3 - and SeO 4 3- showed that the retention fraction would include all species with oxidation state higher or equal to VI. The retentions observed in the different matrices show the existence of four families with respective retentions of: 2.6%[K 2 SeO 4 diluted in (NH 4 ) 2 SO 4 ], 9.2% [anhydrous and hydrated Li and Ca selenates; K 2 SeO 4 diluted in NaIO 3 ; Se + implanted in K 2 SeO 4 ; Triglycine selenate], 21.5% [K 2 SeO 4 diluted in KNO 3 , K 2 SO 4 , Na 2 WO 4 and Na 2 WO 4 .2H 2 O] and 32.0% [Na,K and Cs selenates]. Whereas chemical considerations may be invoked for the (NH 4 ) 2 SO 4 matrix, a mechanical model is proposed for the three other groups. (author)

  2. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  3. Axisymmetric general relativistic hydrodynamics: Long-term evolution of neutron stars and stellar collapse to neutron stars and black holes

    International Nuclear Information System (INIS)

    Shibata, Masaru

    2003-01-01

    We report a new implementation for axisymmetric simulation in full general relativity. In this implementation, the Einstein equations are solved using the Nakamura-Shibata formulation with the so-called cartoon method to impose an axisymmetric boundary condition, and the general relativistic hydrodynamic equations are solved using a high-resolution shock-capturing scheme based on an approximate Riemann solver. As tests, we performed the following simulations: (i) long-term evolution of nonrotating and rapidly rotating neutron stars, (ii) long-term evolution of neutron stars of a high-amplitude damping oscillation accompanied with shock formation, (iii) collapse of unstable neutron stars to black holes, and (iv) stellar collapses to neutron stars. Tests (i)-(iii) were carried out with the Γ-law equation of state, and test (iv) with a more realistic parametric equation of state for high-density matter. We found that this new implementation works very well: It is possible to perform the simulations for stable neutron stars for more than 10 dynamical time scales, to capture strong shocks formed at stellar core collapses, and to accurately compute the mass of black holes formed after the collapse and subsequent accretion. In conclusion, this implementation is robust enough to apply to astrophysical problems such as stellar core collapse of massive stars to a neutron star, and black hole, phase transition of a neutron star to a high-density star, and accretion-induced collapse of a neutron star to a black hole. The result for the first simulation of stellar core collapse to a neutron star started from a realistic initial condition is also presented

  4. Using a Tandem Pelletron accelerator to produce a thermal neutron beam for detector testing purposes.

    Science.gov (United States)

    Irazola, L; Praena, J; Fernández, B; Macías, M; Bedogni, R; Terrón, J A; Sánchez-Nieto, B; Arias de Saavedra, F; Porras, I; Sánchez-Doblado, F

    2016-01-01

    Active thermal neutron detectors are used in a wide range of measuring devices in medicine, industry and research. For many applications, the long-term stability of these devices is crucial, so that very well controlled neutron fields are needed to perform calibrations and repeatability tests. A way to achieve such reference neutron fields, relying on a 3 MV Tandem Pelletron accelerator available at the CNA (Seville, Spain), is reported here. This paper shows thermal neutron field production and reproducibility characteristics over few days. Copyright © 2015 Elsevier Ltd. All rights reserved.

  5. Pilot experimental study on continual spectrum thermal neutron in-line phase contrast radiography

    International Nuclear Information System (INIS)

    Tang Bin; Huo Heyong; Wu Yang

    2009-01-01

    The in-line phase contrast radiography is one of phase contrast imaging methods. The neutron in-line phase contrast is developed with X-rays phase contrast radiography. In the paper, the principle of in-line phase contrast is introduced briefly and the experimental result of thermal neutron in-line contrast at SPRR-300 is analysed. It shows that thermal neutron can be used as in-line phase contrast radiography and enhances the edge of some sample in radiography and complements the disadvantage of conventional neutron radiography. (authors)

  6. Development of the variety for resistance against bacterial leaf-blight in rice with thermal neutrons

    International Nuclear Information System (INIS)

    Nakai, Hirokazu

    1990-01-01

    In search for the development of genes for resistance against bacterial leaf-blight in rice, thermal neutrons generated from the Research Reactor at the Kyoto University have been applied to the breeding. In this paper, the developmental outcome is described, and a potential application of thermal neutrons for breeding the variety of resistance against bacterial leaf-blight in rice is reviewed. When thermal neutrons were delivered to the rice, the ratio of absorbed doses by B-10, which is contained in a small quantity in the plant, was found to be larger than expected. This implies characteristic effects of thermal neutrons on the plant. When boric acid was incorporated into the plant before irradiation, the effect of thermal neutrons per irradiation time was considered to become great. The frequency of mutations for resistance was significantly higher by thermal neutron, as compared with that induced by other mutagens, such as gamma radiation, ethylene-imine, ethyl-methane-sulfonate, and nitroso-methyl-urea. Genetic analysis of mutants for resistance revealed recessive genes and polygenes. Finally, the application of thermal neutrons and other radiations would contribute greatly to a resolution of serious pollution problems in global food and environment. (N.K.)

  7. Determination of thermal neutrons diffusion length in graphite; Determinacion de la Longitud de Difusion de los Neutrones Termicos en Grafito

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Fite, J

    1959-07-01

    The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs.

  8. Neutron polarizing Fe-Al supermirror on a Si crystal substrate and its applications for thermal and cold neutrons

    International Nuclear Information System (INIS)

    Syromyatnikov, V.G.; Shchebetov, A.F.; Soroko, Z.N.

    1994-01-01

    Experimental data are presented for an Fe-Al neutron polarizing supermirror on a Si crystal substrate with an antireflecting Cd layer. The polarizing efficiency of this supermirror is P≥qslant0.8 for the range of glancing angles θ/λ=0.25-1.7 /nm and P≥qslant0.95 for θ/λ=0.34-1.7 /nm. Some applications of this supermirror for thermal and cold neutrons are considered. ((orig.))

  9. IMPROVED COMPUTATIONAL CHARACTERIZATION OF THE THERMAL NEUTRON SOURCE FOR NEUTRON CAPTURE THERAPY RESEARCH AT THE UNIVERSITY OF MISSOURI

    Energy Technology Data Exchange (ETDEWEB)

    Stuart R. Slattery; David W. Nigg; John D. Brockman; M. Frederick Hawthorne

    2010-05-01

    Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. This is essential for detailed dosimetric studies required for the anticipated research program.

  10. A spectral effect on phase evolution in neutron-irradiated aluminum?

    International Nuclear Information System (INIS)

    Farrell, K.

    1994-01-01

    It is proposed that the spatial distribution of transmutation-produced silicon precipitates formed in aluminum by capture of thermal neutrons is systematically modified by the fast neutron portion of the spectrum, and the tensile strength is altered accordingly. The concept is used to rationalize some previously erratic tensile data

  11. Neutronics of rectangular parallelepiped polyethylene moderator in wing geometry for accelerator based thermal neutron source

    International Nuclear Information System (INIS)

    Kiyanagi, Yoshiaki

    1984-01-01

    Numerical and experimental studies of the wing geometry moderator are performed in order to examine (a) the effects of the target position and the moderator thickness on the beam intensity and on the pulse shapes emitted from a polyethylene thermal moderator, and (b) the optimum thickness of the moderator. The beam intensity emitted from the moderator is expressed by an integration of the product of the source neutron distribution and the beam intensity produced by a unit intensity point source in the moderator. By applying this expression mechanism is analyzed for the optimum target position and the saturation phenomena of the intensity and the pulse width emitted from the moderator. The optimum target position is at about 2cm from the neutron emission surface for moderators thicker than 4cm and at about half of the moderator thickness for thinner ones. The intensity and the pulse shapes emitted from the moderator vary little if the target distance is varied around the optimum one and become close to the saturated ones at about 8cm thickness. It is indicated by the analysis of figures of merit that a moderator of 4--6cm thickness is optimum. (author)

  12. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  13. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  14. Influence of orientation averaging on the anisotropy of thermal neutrons scattering on water molecules

    International Nuclear Information System (INIS)

    Markovic, M. I.; Radunovic, J. B.

    1976-01-01

    Determination of spatial distribution of neutron flux in water, most frequently used moderator in thermal reactors, demands microscopic scattering kernels dependence on cosine of thermal neutrons scattering angle when solving the Boltzmann equation. Since spatial orientation of water molecules influences this dependence it is necessary to perform orientation averaging or rotation-vibrational intermediate scattering function for water molecules. The calculations described in this paper and the obtained results showed that methods of orientation averaging do not influence the anisotropy of thermal neutrons scattering on water molecules, but do influence the inelastic scattering

  15. Factors affecting neutron measurements and calculations. Part C. Trace element concentrations in granite and their impact on thermal neutron activation

    International Nuclear Information System (INIS)

    Ruehm, Werner; Huber, Thomas; Nolte, Eckehart; Kato, Kazuo; Imanaka, Tetsuji; Egbert, Stephen D.

    2005-01-01

    Trace elements such as Li, B, Sm, and Gd can, despite their low elemental concentration in mineral materials, influence thermal neutron activation in Hiroshima and Nagasaki samples, due to their high thermal neutron absorption cross sections. This was demonstrated for a granite core, where the addition of those trace elements to the elemental composition of granite reduces the production of 152 Eu by some 25% at a depth of 25 cm from the surface. If typical concentrations of those trace elements are added to DS02 reference soil, however, the production of 152 Eu one meter above ground is not changed significantly, because of the high water content of the soil. This indicates that DS02 soil represents a reasonable reference material for the air-over-ground transport calculations. It must be kept in mind, however, that the local environment of any sample investigated for thermal neutron activation might be characterized by other elemental compositions. In particular, trace element and hydrogen concentrations could be considerably different from those used for DS02 reference soil. As an example it was demonstrated that in a granite gravestone thermal neutron activation of 36 Cl close to the surface might be, in the worst case, reduced by some 30%, due to increased local granite concentration in this type of environment. Beside other parameters such as, for example, individual sample geometry, the variability of trace elements in soil might be one reason for the variability that is observed in the individual thermal neutron activation measurements (Gold 1995). It is necessary, therefore, to carefully model the exposure geometry of the exposed material, its chemical composition, and the surrounding interface materials in order to obtain the best possible agreement in comparisons between calculated and measured data for thermal neutrons. (author)

  16. Enhancement of thermal neutron attenuation of nano-B4C, -BN dispersed neutron shielding polymer nanocomposites

    International Nuclear Information System (INIS)

    Kim, Jaewoo; Lee, Byung-Chul; Uhm, Young Rang; Miller, William H.

    2014-01-01

    Highlights: • Preparation of B 4 C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B 4 C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B 4 C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption

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

    International Nuclear Information System (INIS)

    Ogawa, Yoshihiro

    1985-01-01

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

  18. Enhancement of thermal neutron attenuation of nano-B{sub 4}C, -BN dispersed neutron shielding polymer nanocomposites

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jaewoo, E-mail: kimj@kaeri.re.kr [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); WCI Quantum Beam based Radiation Research Center, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States); Lee, Byung-Chul [Nuclear Reactor Core Design Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Uhm, Young Rang [Radioisotopes Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Miller, William H. [Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States)

    2014-10-15

    Highlights: • Preparation of B{sub 4}C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B{sub 4}C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B{sub 4}C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption.

  19. Neutron stars

    International Nuclear Information System (INIS)

    Irvine, J.M.

    1978-01-01

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

  20. Using TRIGA Mark II research reactor for irradiation with thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

    2015-03-15

    Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

  1. Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Geist, William H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Root, Margaret A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Belian, Anthony P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-02

    The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

  2. Thickness optimization of various moderator materials for maximization of thermal neutron fluence

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

    Plasma focus device is widely being used as pulsed neutron source for variety of applications. Measurements of neutron yield by largely preferred Helium-3 proportional counter and Silver activation counter are mainly sensitive to thermal neutrons and are typically used with a neutron moderator. Thermalization of neutron is based on scattering reaction and hydrogenous materials are the best thermalizing medium. The efficiency of aforementioned neutron detectors is considerably affected by physical and geometrical properties of thermalizing medium i.e. moderator material, its thickness and shape. In view of the same, simulations have been performed to explore the effective utilization of Polyethylene, Perspex and Light water as moderating mediums for cylindrical and spherical geometry. In this study, estimated thermal fluence value up to 0.5 eV has been considered as the benchmark factor for comparing efficient thermalization by specific material, its thickness and shape. In either of the shapes being cylindrical or spherical, use of Polyethylene as moderating medium has resulted in minimum optimum thickness along with highest thermal fluence. (author)

  3. Apparatus and process for continuous measurement of moisture in moving coal by neutron thermalization

    International Nuclear Information System (INIS)

    Stewart, R.F.

    1967-01-01

    The invention relates to an apparatus and process for the measurement of moisture contents in solid materials. More particularly, the invention makes available a continuous moisture analysis of a moving mass of material, such as coal, by penetrating such material with neutrons emitted from a source of fast neutrons and detecting, counting, and recording slowed or thermalized neutrons reflected from the internal structure of the material. (U.S.)

  4. Tables for simplifying calculations of activities produced by thermal neutrons

    Science.gov (United States)

    Senftle, F.E.; Champion, W.R.

    1954-01-01

    The method of calculation described is useful for the types of work of which examples are given. It is also useful in making rapid comparison of the activities that might be expected from several different elements. For instance, suppose it is desired to know which of the three elements, cobalt, nickel, or vanadium is, under similar conditions, activated to the greatest extent by thermal neutrons. If reference is made to a cross-section table only, the values may be misleading unless properly interpreted by a suitable comparison of half-lives and abundances. In this table all the variables have been combined and the desired information can be obtained directly from the values of A 3??, the activity produced per gram per second of irradiation, under the stated conditions. Hence, it is easily seen that, under similar circumstances of irradiation, vanadium is most easily activated even though the cross section of one of the cobalt isotopes is nearly five times that of vanadium and the cross section of one of the nickel isotopes is three times that of vanadium. ?? 1954 Societa?? Italiana di Fisica.

  5. Thermal structure of accreting neutron stars and strange stars

    International Nuclear Information System (INIS)

    Miralda-Escude, J.; Paczynski, B.; Haensel, P.

    1990-01-01

    Steady-state models of accreting neutron stars and strange stars are presented, and their properties as a function of accretion rate are analyzed. The models have steady-state envelopes, with stationary hydrogen burning taken into account, the helium shell flashes artificially suppressed, and the crust with a large number of secondary heat sources. The deep interiors are almost isothermal and are close to thermal equilibrium. A large number of models were calculated for many values of the accretion rates, with ordinary, pion-condensed, and strange cores, with and without secondary heat sources in the crust, and with the heavy element content of the accreting matter in the range Z = 0.0002-0.02. All models show a similar pattern of changes as the accretion rate is varied. For low accretion rates, the hydrogen burning shell is unstable; for intermediate rates, the hydrogen burning shell is stable, but helium burning is not; for high rates, the two shell sources burn together and are unstable. 60 refs

  6. Describing function theory as applied to thermal and neutronic problems

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1983-01-01

    Describing functions have traditionally been used to obtain the solutions of systems of ordinary differential equations. In this work the describing function concept has been extended to include nonlinear, distributed parameter partial differential equations. A three-stage solution algorithm is presented which can be applied to any nonlinear partial differential equation. Two generalized integral transforms were developed as the T-transform for the time domain and the B-transform for the spatial domain. The thermal diffusion describing function (TDDF) is developed for conduction of heat in solids and a general iterative solution along with convergence criteria is presented. The proposed solution method is used to solve the problem of heat transfer in nuclear fuel rods with annular fuel pellets. As a special instance the solid cylindrical fuel pellet is examined. A computer program is written which uses the describing function concept for computing fuel pin temperatures in the radial direction during reactor transients. The second problem investigated was the neutron diffusion equation which is intrinsically different from the first case. Although, for most situations, it can be treated as a linear differential equation, the describing function method is still applicable. A describing function solution is derived for two possible cases: constant diffusion coefficient and variable diffusion coefficient. Two classes of describing functions are defined for each case which portray the leakage and absorption phenomena. For the specific case of a slab reactor criticality problem the comparison between analytical and describing function solutions revealed an excellent agreement

  7. Thermal neutron detection by activation of CaSO4:Dy + KBr thermoluminescent phosphors

    International Nuclear Information System (INIS)

    Gordon, A.M.P.L.; Muccillo, R.

    1979-01-01

    Thermoluminescence (TL) studies to detect thermal neutrons were performed in cold-pressed CaSO 4 :0,1%Dy + KBr samples. The detection is based on the self-irradiation of the CaSO 4 :Dy TL phosphor by the Br isotopes activated by exposure to a mixed neutron-gamma field. (Author) [pt

  8. Measurement and simulation of thermal neutron flux distribution in the RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

    2018-01-01

    The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

  9. Thermal neutron cross sections and resonance integrals for the 1994 handbook of chemistry and physics

    International Nuclear Information System (INIS)

    Holden, N.E.

    1994-01-01

    A re-evaluation of all thermal neutron cross sections and neutron resonance integrals has been performed, utilizing the previous database of the ''Barn Book'' and all of the more recently published experiments. Only significant changes or previously undetermined values are recorded in this report. The source for each value is also recorded in the accompanying table

  10. Measurements of thermal and fast neutron fluxes at the TRIGA reactor

    International Nuclear Information System (INIS)

    Zerdin, F.; Grabovsek, Z.; Klinc, T.; Solinc, H.

    1966-01-01

    Gold foils were placed at different positions in the TRIGA reactor core and in the experimental devices. Absolute values of the thermal neutron flux at these positions were obtained by coincidence method. Preliminary fast neutron spectrum was measured by threshold detector and by 'Li 6 sandwich' detector. A short description of the applied method and obtained measurements results are included [sl

  11. Parity non-conserving effects in thermal neutron-deuteron radiative capture

    International Nuclear Information System (INIS)

    Desplanques, B.

    1985-01-01

    Predictions of parity non-conserving effects in thermal neutron-deuteron radiative capture are presented. The sensitivity of the results to models of the strong interaction as well as the validity of approximations made in previous calculations are discussed

  12. Impact of thermal and intermediate energy neutrons on the semiconductor memories for the CERN accelerators

    CERN Document Server

    Cecchetto, Matteo; Gerardin, Simone

    A wide quantity of SRAM memories are employed along the Large Hadron Collider (LHC), the main CERN accelerator, and they are subjected to high levels of ionizing radiations which compromise the reliability of these devices. The Single Event Effect (SEE) qualification for components to be used in the complex high-energy accelerator at CERN relies on the characterization of two cross sections: 200-MeV protons and thermal neutrons. However, due to cost and time constraints, it is not always possible to characterize the SEE response of components to thermal neutrons, which is often regarded as negligible for components without borophosphosilicate glass (BPSG). Nevertheless, as recent studies show, the sensitivity of deep sub-micron technologies to thermal neutrons has increased owing to the presence of Boron 10 as a dopant and contact contaminant. The very large thermal neutron fluxes relative to high-energy hadron fluxes in some of the heavily shielded accelerator areas imply that even comparatively small therm...

  13. Computational features of the MELT-III neutronics, thermal-hydraulics computer code system

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Waltar, A.E.

    1976-01-01

    A multichannel, thermal-hydraulics, neutronic accident analysis program for simulating fast reactor behavior from a hypothetical accident inception to the start of core disassembly or to reactor shutdown is described

  14. Identification of the fast and thermal neutron characteristics of transuranic waste drums

    Energy Technology Data Exchange (ETDEWEB)

    Storm, B.H. Jr.; Bramblett, R.L. [Lockheed Martin Specialty Components, Largo, FL (United States); Hensley, C. [Oak Ridge National Lab., TN (United States)

    1997-11-01

    Fissile and spontaneously fissioning material in transuranic waste drums can be most sensitively assayed using an active and passive neutron assay system such as the Active Passive Neutron Examination and Assay. Both the active and the passive assays are distorted by the presence of the waste matrix and containerization. For accurate assaying, this distortion must be characterized and accounted for. An External Matrix Probe technique has been developed that accomplishes this task. Correlations between in-drum neutron flux measurements and monitors in the Active Passive Neutron Examination and Assay chamber with various matrix materials provide a non-invasive means of predicting the thermal neutron flux in waste drums. Similarly, measures of the transmission of fast neutrons emitted from sources in the drum. Results obtained using the Lockheed Martin Specialty Components Active Passive Neutron Examination and Assay system are discussed. 12 figs., 1 tab.

  15. Determination of cadmium in zinc ores by thermal neutron absorption analysis

    International Nuclear Information System (INIS)

    De Norre, L.; Op de Beeck, J.; Hoste, J.

    1983-01-01

    A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux in related to the 52 V activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constituents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci 227 Ac-Be isotope neutron source. After a decay of 30 s, the 52 V activity of the vanadium detector is measured for 400 s with a NaI(Tl) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI-11 Microprocessor with 12Kx16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54%, resp. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  17. Thermal neutron inelastic scattering and it's application to the material science

    International Nuclear Information System (INIS)

    Li Zhuqi

    1986-01-01

    A brief description of the elementary scattering theory of the interaction between the thermal neutrons and the condensed matter is given and the characteristics related to the experimental method of the thermal neutrons inelastic scattering is described. Expressions of the phonons dispersion, density of the phonon state and the self-diffusion coefficient at the some conditions are also introduced. Some examples of describing diagram of the phonon dispersion, density of the phonons state and selfdiffusion coefficient measured by different authors are given

  18. TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

    International Nuclear Information System (INIS)

    Gowins, G.

    1984-01-01

    Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

  19. Thermal states of coldest and hottest neutron stars in soft X-ray transients

    OpenAIRE

    Yakovlev, D. G.; Levenfish, K. P.; Potekhin, A. Y.; Gnedin, O. Y.; Chabrier, G.

    2003-01-01

    We calculate the thermal structure and quiescent thermal luminosity of accreting neutron stars (warmed by deep crustal heating in accreted matter) in soft X-ray transients (SXTs). We consider neutron stars with nucleon and hyperon cores and with accreted envelopes. It is assumed that an envelope has an outer helium layer (of variable depth) and deeper layers of heavier elements, either with iron or with much heavier nuclei (of atomic weight A > 100) on the top (Haensel & Zdunik 1990, 2003, as...

  20. Ohmic ion temperature and thermal diffusivity profiles from the JET neutron emission profile monitor

    Energy Technology Data Exchange (ETDEWEB)

    Esposito, B. (ENEA, Frascati (Italy). Centro Ricerche Energia); Marcus, F.B.; Conroy, S.; Jarvis, O.N.; Loughlin, M.J.; Sadler, G.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell (United Kingdom))

    1993-10-01

    The JET neutron emission profile monitor was used to study ohmically heated deuterium discharges. The radial profile of the neutron emissivity is deduced from the line-integral data. The profiles of ion temperature, T[sub i], and ion thermal diffusivity, [chi][sub i], are derived under steady-state conditions. The ion thermal diffusivity is higher than, and its scaling with plasma current opposite to, that predicted by neoclassical theory. (author).

  1. Ohmic ion temperature and thermal diffusivity profiles from the JET neutron emission profile monitor

    International Nuclear Information System (INIS)

    Esposito, B.

    1993-01-01

    The JET neutron emission profile monitor was used to study ohmically heated deuterium discharges. The radial profile of the neutron emissivity is deduced from the line-integral data. The profiles of ion temperature, T i , and ion thermal diffusivity, χ i , are derived under steady-state conditions. The ion thermal diffusivity is higher than, and its scaling with plasma current opposite to, that predicted by neoclassical theory. (author)

  2. Experimental characterization of HOTNES: A new thermal neutron facility with large homogeneity area

    Energy Technology Data Exchange (ETDEWEB)

    Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Sperduti, A. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pietropaolo, A.; Pillon, M. [ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN–Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2017-01-21

    A new thermal neutron irradiation facility, called HOTNES (HOmogeneous Thermal NEutron Source), was established in the framework of a collaboration between INFN-LNF and ENEA-Frascati. HOTNES is a polyethylene assembly, with about 70 cmx70 cm square section and 100 cm height, including a large, cylindrical cavity with diameter 30 cm and height 70 cm. The facility is supplied by a {sup 241}Am-B source located at the bottom of this cavity. The facility was designed in such a way that the iso-thermal-fluence surfaces, characterizing the irradiation volume, coincide with planes parallel to the cavity bottom. The thermal fluence rate across a given isofluence plane is as uniform as 1% on a disk with 30 cm diameter. Thermal fluence rate values from about 700 cm{sup −2} s{sup −1} to 1000 cm{sup −2} s{sup −1} can be achieved. The facility design, previously optimized by Monte Carlo simulation, was experimentally verified. The following techniques were used: gold activation foils to assess the thermal fluence rate, semiconductor-based active detector for mapping the irradiation volume, and Bonner Sphere Spectrometer to determine the complete neutron spectrum. HOTNES is expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

  4. Simultaneous measurement of fission fragments and prompt neutrons for thermal neutron-induced fission of U-235

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Katsuhisa; Yamamoto, Hideki; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan)

    1997-03-01

    Simultaneous measurement of fission fragments and prompt neutrons following the thermal neutron induced fission of U-235 has been performed in order to obtain the neutron multiplicity (v) and its emission energy ({eta}) against the specified mass (m{sup *}) and the total kinetic energy (TKE). The obtained value of -dv/dTKE(m{sup *}) showed a saw-tooth distribution. The average neutron energy <{eta}>(m{sup *}) had a distribution with a reflection symmetry around the half mass division. The measurement also gave the level density parameters of the specified fragment, a(m{sup *}), and this parameters showed a saw-tooth trend too. The analysis by a phenomenological description of this parameters including the shell and collective effects suggested the existence of a collective motion of the fission fragments. (author)

  5. Evaluating the 239Pu Prompt Fission Neutron Spectrum Induced by Thermal to 30 MeV Neutrons

    Directory of Open Access Journals (Sweden)

    Neudecker D.

    2016-01-01

    Full Text Available We present a new evaluation of the 239Pu prompt fission neutron spectrum (PFNS induced by thermal to 30 MeV neutrons. Compared to the ENDF/B-VII.1 evaluation, this one includes recently published experimental data as well as an improved and extended model description to predict PFNS. For instance, the pre-equilibrium neutron emission component to the PFNS is considered and the incident energy dependence of model parameters is parametrized more realistically. Experimental and model parameter uncertainties and covariances are estimated in detail. Also, evaluated covariances are provided between all PFNS at different incident neutron energies. Selected evaluation results and first benchmark calculations using this evaluation are briefly discussed.

  6. A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-08-01

    The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent

  7. Indirect and direct measurement of thermal neutron acceleration by inelastic scattering on the 177Lu isomer

    International Nuclear Information System (INIS)

    Belier, G.; Roig, O.; Meot, V.; Daugas, J.M.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Veyssiere, Ch.

    2008-01-01

    When neutrons interact with isomers, these isomers can de-excite and in such a reaction the outgoing neutron has an energy greater than the in-going one. This process is referred as Inelastic Neutron Acceleration or Super-elastic Scattering. Up to now this process was observed for only two nucleus, 152m Eu and 180m Hf by measuring the number of fast neutrons produced by isomeric targets irradiated with thermal neutrons. In these experiments the energies of the accelerated neutrons were not measured. This report presents an indirect measurement of inelastic neutron acceleration on 177m Lu, based on the burn-up and the radiative capture cross sections measurements. Since at thermal energies the inelastic scattering and the radiative capture are the only processes that contribute to the isomer burn-up, the inelastic cross section can be deduced from the difference between the two measured quantities. Applying this method for the 177 Lu isomer with different neutron fluxes we obtained a value of (257 ± 50) barns (for a temperature of 323 K) and determined that there is no integral resonance for this process. In addition the radiative capture cross section on 177g Lu was measured with a much better accuracy than the accepted value. Since the acceleration cross section is quite high, a direct measurement of this process was undertaken, sending thermal neutrons and measuring the fast neutrons. The main goal now is to measure the outgoing neutron energies in order to identify the neutron transitions in the exit channel. In particular the K conservation question can be addressed by such a measurement. (author)

  8. Determination of Thermal Neutron Capture Cross Sections Using Cold Neutron Beams at the Budapest PGAA-NIPS Facilities

    International Nuclear Information System (INIS)

    Belgya, T.

    2006-01-01

    A complete elemental gamma-ray library was measured with our guided thermal beam at the Budapest PGAA facility in the period of 1995-2000. Using this data library in an IAEA CRP on PGAA it was managed to re-normalize the ENSDF intensity data with the Budapest intensities. Based on this renormalization thermal neutron cross sections were deduced for several isotopes. Most of these calculations were done by Richard B. Firestone. The Budapest PGAA-NIPS facilities have been used for routine prompt gamma activation analysis with cold neutrons since the year of 2000. The advantage of the cold neutron beam is that the neutron guide has much higher neutron transmission. This resulted in a gain factor about 20 relative to our thermal guide. For the analytical works a precise comparator technique was developed that is routinely used to determine partial gamma-ray production cross sections. An additional development of our methodology was necessary to be worked out to determine thermal neutron capture cross sections based on the partial gamma-ray production cross sections. In this talk our methodology of radiative capture cross section determination will be presented, including our latest results on 129 I, 204,206,207 Pb and 209 Bi. Most of these works were done in cooperation with people from EU-JRC-IRMM, Geel, Belgium and CEA Cadarache, France. Many partial cross sections of short lived nuclei have been re-measured with our new chopper technique. The uncertainty calculations of the radiative capture cross section determination procedures will be also shown. (authors)

  9. Influence of impurities on the evolution of vacancy-type defects in neutron-irradiated nickel

    International Nuclear Information System (INIS)

    Druzhkov, A.P.; Perminov, D.A.; Arbuzov, V.L.

    2012-01-01

    Highlights: ► We study, by means of PAS, the effects of purity on damage evolution in neutron-irradiated Ni at 330 K. ► Impurity carbon atoms in solution decrease the cascade efficiency during irradiation. ► C–V complexes are formed on the recovery stage III in impure Ni irradiated with 10 −4 dpa. ► The formation of V-loops and SFTs dominate on stage III with increasing dose level. ► The thermal stability of SFTs in impure Ni is similar to that in pure Ni. - Abstract: In order to investigate the effect of impurities on vacancy defect evolution in nickel, specimens with high (5N) and technical (3N) purity were neutron-irradiated at ∼330 K in the IVV-2M reactor (Russia) to fluencies in the range of 1 × 10 21 –1 × 10 23 n/m 2 (E > 0.1 MeV) corresponding to displacement dose levels in the range of about 0.0001–0.01 dpa and subsequently stepwise annealed to about 900 K. The specimens of Ni with different purities were characterized both in as-irradiated state as well as after post-irradiation annealing by positron annihilation spectroscopy. The formation of three-dimensional vacancy clusters (3D-VCs) in cascades was observed under neutron irradiation. The density and size of 3D-VCs depended not only on dose level, but also on purity. The population of 3D-VCs in the technical Ni is lower than that in the high-purity Ni. 3D-VCs collapse into secondary-type clusters (stacking fault tetrahedra (SFTs) and vacancy loops) during stepwise annealing at 350–450 K (stage III in Ni). The suppression of secondary cluster formation in 3N Ni is attributed to an effective vacancy interaction with impurity carbon atoms, which based on a relatively large vacancy–carbon atom binding energy (0.32–0.35 eV). The trapping of vacancies released at the collapse of 3D-VCs by the interstitial impurity atoms dominates at low irradiation dose level (10 −4 dpa). Thus, we found that carbon impurity atoms have strong effects both on the primary vacancy-type defect

  10. Thermal neutron diffusion parameters dependent on the flux energy distribution in finite hydrogenous media

    International Nuclear Information System (INIS)

    Drozdowicz, K.

    1999-01-01

    Macroscopic parameters for a description of the thermal neutron transport in finite volumes are considered. A very good correspondence between the theoretical and experimental parameters of hydrogenous media is attained. Thermal neutrons in the medium possess an energy distribution, which is dependent on the size (characterized by the geometric buckling) and on the neutron transport properties of the medium. In a hydrogenous material the thermal neutron transport is dominated by the scattering cross section which is strongly dependent on energy. A monoenergetic treatment of the thermal neutron group (admissible for other materials) leads in this case to a discrepancy between theoretical and experimental results. In the present paper the theoretical definitions of the pulsed thermal neutron parameters (the absorption rate, the diffusion coefficient, and the diffusion cooling coefficient) are based on Nelkin's analysis of the decay of a neutron pulse. Problems of the experimental determination of these parameters for a hydrogenous medium are discussed. A theoretical calculation of the pulsed parameters requires knowledge of the scattering kernel. For thermal neutrons it is individual for each hydrogenous material because neutron scattering on hydrogen nuclei bound in a molecule is affected by the molecular dynamics (characterized with internal energy modes which are comparable to the incident neutron energy). Granada's synthetic model for slow-neutron scattering is used. The complete up-dated formalism of calculation of the energy transfer scattering kernel after this model is presented in the paper. An influence of some minor variants within the model on the calculated differential and integral neutron parameters is shown. The theoretical energy-dependent scattering cross section (of Plexiglas) is compared to experimental results. A particular attention is paid to the calculation of the diffusion cooling coefficient. A solution of an equation, which determines the

  11. Dosimeter incorporating radiophotoluminescent detectors for thermal neutrons and γ-rays in n-γ fields

    Energy Technology Data Exchange (ETDEWEB)

    Salem, Y.O. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France); Nachab, A., E-mail: a.nachab@uca.ma [Département de physique, Faculté Poly-disciplinaire, Université Cadi Ayyad, Route Sidi Bouzid BP 4162, 46000 Safi (Morocco); Roy, C.; Nourreddine, A. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France)

    2016-10-15

    We have developed a dosimeter associating different neutron converters with two radiophotoluminescent detectors to measure thermal neutrons and γ-rays in a mixed n-γ field. Tests show that the H{sup ∗}(10) and H{sub p}(10) responses to thermal neutrons and γ-rays are linear with detection limits lower than 0.4 mSv. The angular dependence of the dosimeter response is satisfactory and the influence of a phantom on the results is examined.

  12. Detection mechanisms in silicon diodes used as α-particle and thermal neutron detectors

    International Nuclear Information System (INIS)

    Cerofolini, G.F.; Ferla, G.; Foglio Para, A.

    1981-01-01

    Some common silicon devices (diodes, RAMs etc.) can be used as α and thermal neutron detectors. An α resolution of approx. equal to 3% can be obtained utilizing p + /n or n + /p diodes with no external bias. Thermal neutrons are detected by means of the reaction 10 B(n,α) 7 Li on the 10 B present in the devices. Neutron efficiency has been substantially improved by implantation of 10 B ions in the p + region of the diodes. Experimental results allow us to clarify the carrier collection mechanisms throughout the device. Some current opinions in the field are contradicted. (orig.)

  13. Studies on thermal neutron perturbation factor needed for bulk sample activation analysis

    CERN Document Server

    Csikai, J; Sanami, T; Michikawa, T

    2002-01-01

    The spatial distribution of thermal neutrons produced by an Am-Be source in a graphite pile was measured via the activation foil method. The results obtained agree well with calculated data using the MCNP-4B code. A previous method used for the determination of the average neutron flux within thin absorbing samples has been improved and extended for a graphite moderator. A procedure developed for the determination of the flux perturbation factor renders the thermal neutron activation analysis of bulky samples of unknown composition possible both in hydrogenous and graphite moderators.

  14. Measurement of the diffusion length of thermal neutrons in the beryllium oxide

    International Nuclear Information System (INIS)

    Koechlin, J.C.; Martelly, J.; Duggal, V.P.

    1955-01-01

    The diffusion length of thermal neutrons in the beryllium oxide has been obtained while studying the spatial distribution of the neutrons in a massive parallelepiped of this matter placed before the thermal column of the reactor core of Saclay. The mean density of the beryllium oxide (BeO) is 2,95 gr/cm 3 , the mean density of the massif is 2,92 gr/cm 3 . The value of the diffusion length, deducted of the done measures, is: L = 32,7 ± 0,5 cm (likely gap). Some remarks are formulated about the influence of the spectral distribution of the neutrons flux used. (authors) [fr

  15. Thermal neutron pulsed parameters in non-hydrogenous systems. Experiment for lead grains

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Gabanska, B.; Kosik, M.; Krynicka, E.; Woznicka, U.; Zaleski, T.

    1997-01-01

    In Czubek's method of measurement of the thermal neutron macroscopic absorption cross section a two-region geometry is applied where the investigated sample is surrounded by an external moderator. Both regions in the measurements made up till now were hydrogenous, which means the same type of the thermal neutron transport properties. In the paper a theoretical consideration to use non-hydrogenous materials as the samples is presented. Pulsed neutron measurements have been performed on homogeneous material in a geometry of the classic experiment with the variable geometric buckling. Two decay constants have been measured for different cylindrical samples of small lead grains (a lead shot). (author)

  16. The Dynamic Method for Time-of-Flight Measurement of Thermal Neutron Spectra from Pulsed Sources

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tulaev, A.B.; Bobrakov, V.F.

    1994-01-01

    The time-of-flight method for a measurement of thermal neutron spectra in the pulsed neutron sources with high efficiency of neutron registration, more than 10 5 times higher in comparison with traditional one, is described. The main problems connected with the electric current technique for time-of-flight spectra measurement are examined. The methodical errors, problems of a special neutron detector design and other questions are discussed. Some experimental results, spectra from surfaces of the water and solid methane moderators, obtained in the pulsed reactor IBR-2 (Dubna, Russia) are presented. 4 refs., 5 figs

  17. Thermal neutron capture cross section for the K isomer 177Lum

    International Nuclear Information System (INIS)

    Belier, G.; Roig, O.; Daugas, J.-M.; Giarmana, O.; Meot, V.; Letourneau, A.; Marie, F.; Foucher, Y.; Aupiais, J.; Abt, D.; Jutier, Ch.; Le Petit, G.; Bettoni, C.; Gaudry, A.; Veyssiere, Ch.; Barat, E.; Dautremer, T.; Trama, J.-Ch.

    2006-01-01

    The thermal neutron radiative capture cross section for the K isomeric state in 177 Lu has been measured for the first time. Several 177 Lu m targets have been prepared and irradiated in various neutron fluxes at the Lauee Langevin Institute in Grenoble and at the CEA reactors OSIRIS and ORPHEE in Saclay. The method consists of measuring the 178 Lu activity by γ-ray spectroscopy. The values obtained in four different neutron spectra have been used to calculate the resonance integral of the radiative capture cross section for 177 Lu m . In addition, an indirect method leads to the determination of the 177 Lu g neutron radiative capture cross section

  18. Search for sp-interference effect in emission of prompt neutrons of sup 2 sup 3 sup 5 U fission by thermal polarized neutrons

    CERN Document Server

    Danilyan, G V; Pavlov, V S; Fedorov, A V

    2001-01-01

    The results of the experiment for the search of the sp-interference effect in the distribution of the prompt neutrons of the sup 2 sup 3 sup 5 U fission by thermal polarized neutrons are presented. The experiment is carried out on the polarized neutrons beam of the MIFI reactor. The scheme of the installation and the flight time spectrum are presented

  19. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    International Nuclear Information System (INIS)

    Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  20. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  1. Moduli evolution in the presence of thermal corrections

    International Nuclear Information System (INIS)

    Barreiro, Tiago; Carlos, Beatriz de; Copeland, Edmund J.; Nunes, Nelson J.

    2008-01-01

    We study the effect of thermal corrections on the evolution of moduli in effective supergravity models. This is motivated by previous results in the literature suggesting that these corrections could alter and even erase the presence of a minimum in the zero temperature potential, something that would have disastrous consequences in these particular models. We show that, in a representative sample of flux compactification constructions, this need not be the case, although we find that the inclusion of thermal corrections can dramatically decrease the region of initial conditions for which the moduli are stabilized. Moreover, the bounds on the reheating temperature coming from demanding that the full, finite temperature potential, has a minimum can be considerably relaxed given the slow pace at which the evolution proceeds.

  2. Thermal neutron radiative capture cross-section of 186W(n, γ)187W reaction

    International Nuclear Information System (INIS)

    Tan, V H; Son, P N

    2016-01-01

    The thermal neutron radiative capture cross section for 186 W(n, γ) 187 W reaction was measured by the activation method using the filtered neutron beam at the Dalat research reactor. An optimal composition of Si and Bi, in single crystal form, has been used as neutron filters to create the high-purity filtered neutron beam with Cadmium ratio of R cd = 420 and peak energy E n = 0.025 eV. The induced activities in the irradiated samples were measured by a high resolution HPGe digital gamma-ray spectrometer. The present result of cross section has been determined relatively to the reference value of the standard reaction 197 Au(n, γ) 198 Au. The necessary correction factors for gamma-ray true coincidence summing, and thermal neutron self-shielding effects were taken into account in this experiment by Monte Carlo simulations. (paper)

  3. Formation and Evolution of Neutron Star Binaries: Masses of Neutron Stars

    Directory of Open Access Journals (Sweden)

    Lee Chang-Hwan

    2012-02-01

    Full Text Available Neutron star (NS is one of the most interesting astrophysical compact objects for hardronic physics. It is believed that the central density of NS can reach several times the normal nuclear matter density (ρ0. Hence, the inner part of NS is the ultimate testing place for the physics of dense matter. Recently, the mass of NS in a NS-white dwarf (WD binary PSR J1614-2230 has been estimated to be 1.97 ± 0.04M๏ [1]. Since this estimate is based on the observed Shapiro delay, it can give the lower limit of the maximum NS mass and rules out many soft equations of state. On the other hand, all the well-measured NS masses in NS-NS binaries are smaller than 1.5M๏. In this work, by introducing the supercritical accretion during the binary evolution, we propose a possibility of forming higher mass NS in NS-WD binaries. In this scenario, the lifetimes of NS and WD progenitors are significantly different, and NS in NS-WD binary can accrete > 0.5M๏ after NS formation during the giant phase of the progenitor of WD. On the other hand, for the binary system with NS and heavier (> 8M๏ giants, the first-born NS will accrete more from the companion and can collapse into black hole. The only way to avoid the supercritical accretion is that the initial masses of progenitors of NS binary should be very close so that they evolve almost at the same time and don’t have time to accrete after NS formation.

  4. Superconducting Strips: A Concept in Thermal Neutron Detection

    Directory of Open Access Journals (Sweden)

    Vittorio Merlo

    2018-03-01

    Full Text Available In the never-ending quest for better detection efficiency and spatial resolution, various thermal neutron detection schemes have been proposed over the years. Given the presence of some converting layers (typically boron, but 6LiF is also widely used nowadays, the shift towards concepts based on solid state detectors has been steadily increasing and ingenious schemes thereby proposed. However, a trade-off has been always sought for between efficiency and spatial resolution; the problem can be (at least partially circumvented using more elaborate geometries, but this complicates the sample preparation and detector construction. Thus, viable alternatives must be found. What we proposed (and verified experimentally is a detection scheme based on the superconducting to normal transition. More precisely, using a boron converting layer, the α particles (generated in the (n, α reaction crossing a low critical temperature superconducting strip some 10 µm wide have been detected; the process, bolometric in nature and based on the ionization energy loss, is intrinsically fast and the spatial resolution very appealing. In this work, some of the work done so far will be illustrated, together with the principles of the measurement and various related problems. The realization of the detector is based on industrial deposition and photolitographic techniques well within the grasp of a condensed matter laboratory, so that there is substantial room for improvement over our elementary strip geometry. Some of the plans for future work will also be presented, together with some improvements both in the choice of the materials and the geometry of the detector.

  5. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    International Nuclear Information System (INIS)

    Hawari, Ayman

    2014-01-01

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  6. The monostandard method in thermal neutron activation analysis of geological, biological and environmental materials

    International Nuclear Information System (INIS)

    Alian, A.; Djingova, R.G.; Kroener, B.; Sansoni, B.

    1984-01-01

    A simple method is described for instrumental multielement thermal neutron activation analysis using a monostandard. For geological and air dust samples, iron is used as a comparator, while sodium has advantages for biological materials. To test the capabilities of this method, the values of the effective cross sections of the 23 elements determined were evaluated in a reactor site with an almost pure thermal neutron flux of about 9x10 12 nxcm -2 xs -1 and an epithermal neutron contribution of less than 0.03%. The values obtained were found to agree mostly well with the best literature values of thermal neutron cross sections. The results of an analysis by activation in the same site agree well with the relative method using multielement standards and for several standard reference materials with certified element contents. A comparison of the element contents obtained by the monostandard and relative methods together with corresponding precisions and accuracies is given. (orig.) [de

  7. Structural evolution of tunneling oxide passivating contact upon thermal annealing.

    Science.gov (United States)

    Choi, Sungjin; Min, Kwan Hong; Jeong, Myeong Sang; Lee, Jeong In; Kang, Min Gu; Song, Hee-Eun; Kang, Yoonmook; Lee, Hae-Seok; Kim, Donghwan; Kim, Ka-Hyun

    2017-10-16

    We report on the structural evolution of tunneling oxide passivating contact (TOPCon) for high efficient solar cells upon thermal annealing. The evolution of doped hydrogenated amorphous silicon (a-Si:H) into polycrystalline-silicon (poly-Si) by thermal annealing was accompanied with significant structural changes. Annealing at 600 °C for one minute introduced an increase in the implied open circuit voltage (V oc ) due to the hydrogen motion, but the implied V oc decreased again at 600 °C for five minutes. At annealing temperature above 800 °C, a-Si:H crystallized and formed poly-Si and thickness of tunneling oxide slightly decreased. The thickness of the interface tunneling oxide gradually decreased and the pinholes are formed through the tunneling oxide at a higher annealing temperature up to 1000 °C, which introduced the deteriorated carrier selectivity of the TOPCon structure. Our results indicate a correlation between the structural evolution of the TOPCon passivating contact and its passivation property at different stages of structural transition from the a-Si:H to the poly-Si as well as changes in the thickness profile of the tunneling oxide upon thermal annealing. Our result suggests that there is an optimum thickness of the tunneling oxide for passivating electron contact, in a range between 1.2 to 1.5 nm.

  8. Nondestructive elemental analysis of coins using accelerator-based thermal neutrons

    International Nuclear Information System (INIS)

    Khairi, F.Z.; Aksoy, A.; Al-Haddad, M.N.

    2007-01-01

    The accelerator-based thermal-neutrons activation analysis setup at KFUPM has an adequate thermal -neutron flux that can be advantageously used for the elemental analysis of a variety of samples including archeological ones. The thermal neutrons are derived from the moderation of fast neutrons from the D (d, n) He reaction which produces fast 2.5 MeV neutrons. A maximum thermals flux of about 2.5x10 n/m-s was achieved. For the purpose of determining the suitability of the set up for the analysis of contemporary and ancient coins, we carried out a feasibility study by irradiating a selected number of Saudi Arabian coins dating from 1958 to 1987 in the thermal-neutron flux. The induced gamma-ray activities were then counted using a HP-GMX detector coupled to a PC-based data acquisition and analysis system. The elements that were determined in the coins were copper (75%), nickel (around 25%) and manganese (<0.5%). Calibration curves were also established for these elements. The determined concentrations are in agreement with the data published by the Standard Catalogue of World Coins. (author)

  9. Study on thermal neutron spectra in reactor moderators by time-of-flight method

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-12-01

    Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

  10. Geant4 Analysis of a Thermal Neutron Real-Time Imaging System

    Science.gov (United States)

    Datta, Arka; Hawari, Ayman I.

    2017-07-01

    Thermal neutron imaging is a technique for nondestructive testing providing complementary information to X-ray imaging for a wide range of applications in science and engineering. Advancement of electronic imaging systems makes it possible to obtain neutron radiographs in real time. This method requires a scintillator to convert neutrons to optical photons and a charge-coupled device (CCD) camera to detect those photons. Alongside, a well collimated beam which reduces geometrical blurriness, the use of a thin scintillator can improve the spatial resolution significantly. A representative scintillator that has been applied widely for thermal neutron imaging is 6LiF:ZnS (Ag). In this paper, a multiphysics simulation approach for designing thermal neutron imaging system is investigated. The Geant4 code is used to investigate the performance of a thermal neutron imaging system starting with a neutron source and including the production of charged particles and optical photons in the scintillator and their transport for image formation in the detector. The simulation geometry includes the neutron beam collimator and sapphire filter. The 6LiF:ZnS (Ag) scintillator is modeled along with a pixelated detector for image recording. The spatial resolution of the system was obtained as the thickness of the scintillator screen was varied between 50 and 400 μm. The results of the simulation were compared to experimental results, including measurements performed using the PULSTAR nuclear reactor imaging beam, showing good agreement. Using the established model, further examination showed that the resolution contribution of the scintillator screen is correlated with its thickness and the range of the neutron absorption reaction products (i.e., the alpha and triton particles). Consequently, thinner screens exhibit improved spatial resolution. However, this will compromise detection efficiency due to the reduced probability of neutron absorption.

  11. Optimized sub thermal neutron source to Linac of CAB

    International Nuclear Information System (INIS)

    Torres, L; Granada, R

    2006-01-01

    We present the results of calculations performed with the code M C N P relative to the neutron field behavior within the moderator for the Bariloche-Linac cold neutron source, using polyethylene as pre moderator and solid mesitylene as moderating material at 90 K.The optimum dimensions for a moderator were obtained, with and without a pre moderator, from the point of view of neutron production and time-width of the neutron pulse.Finally, we adopted for our cold neutron source, a slab pre moderator of P L E at room temperature, and a cylindrical moderator of mesitylene at 90 K with a cooler system of stainless steel with windows of Zircaloy-4 [es

  12. Interaction of thermal and cold neutrons with solids

    International Nuclear Information System (INIS)

    Kilany, M.M.A.

    1986-01-01

    The present thesis deals with total neutron cross-section measurements carried out for germanium - single crystal in the energy range from 2.2 eV to 2.5 MeV, at liquid nitrogen temperature (80 K), room temperature and (440 ± 3) K. Moreover, it includes the transmitted reactor spectrum through the Ge - single crystal with different orientations w.r.t. the neutron beam direction. This thesis also deals with the cross - section measurements of polycrystalline graphite in the energy range from 0.5 eV to 1.3 MeV (neutron wavelength from 0.4 A to 7.8 A). The work also presents the neutron transmission measurements of pyrolytic graphite (P.G) crystal in a neutron wavelength band from 0.3 A to 5.0 A , at different orientations of the crystal w.r.t. the beam direction

  13. Mantle differentiation and thermal evolution of Mars, Mercury, and Venus

    International Nuclear Information System (INIS)

    Spohn, T.

    1991-01-01

    In the present models for the thermal evolution of Mercury, Venus, and Mars encompass core and mantle chemical differentiation, lithospheric growth, and volcanic heat-transfer processes. Calculation results indicate that crust and lithosphere thicknesses are primarily dependent on planet size as well as the bulk concentration of planetary radiogenic elements and the lithosphere's thermal conductivity. The evidence for Martian volcanism for at least 3.5 Gyr, and in Mercury for up to 1 Gyr, in conjunction with the presence of a magnetic field on Mercury and its absence on Mars, suggest the dominance of a lithospheric conduction heat-transfer mechanism in these planets for most of their thermal history; by contrast, volcanic heat piping may have been an important heat-transfer mechanism on Venus. 50 refs

  14. Evolution of a blue supergiant with a neutron star companion immersed in its envelope

    International Nuclear Information System (INIS)

    Delgado, A.J.

    1980-01-01

    The evolution of a binary system consisting of 1 Msub(sun) neutron star and a 25 Msub(sun) blue supergiant through a phase of common envelope is investigated. We include the effects of an additional energy source on the supergiant's envelope, due to the presence of the neutron star, and variable mass loss from the system, taken as proportional to the total luminosity. The results indicate that, independently of the initial period, the system loses its whole envelope as a consequence of the common envelope phase, the final product of this being a detached system, consisting of a neutron star and a helium star. (orig.)

  15. Measure of the slowing down area of thermal state 1,45 eV neutrons in the graphite; Mesure de l'aire de ralentissement des neutrons dans le graphite de 1,45 eV a l'etat thermique

    Energy Technology Data Exchange (ETDEWEB)

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

    1956-12-15

    The life of a neutron emitted as a fast one within a moderating medium may be divided into two parts; at first it is slowed down colliding with the nuclei of the medium; later on, as soon as its energy is of the order of the thermal agitation energy, it diffuses until captured by some nucleus. As a matter of fact, dividing the evolution of a neutron into two phases, a slowing down phase and a diffusing phase is arbitrary, since the thermal equilibrium is attained stepwise. This report is intended to examine the means of inferring the distribution of thermal neutrons from the distribution of fast neutrons on the basis of the distribution of nascent thermal neutrons and of the law of scattering. (author) [French] La vie d'un neutron emis a l'etat rapide dans un milieu moderateur comprend 2 parties: au debut il se ralentit par chocs sur les noyaux du milieu; ensuite, lorsque son energie est de l'ordre de l'energie d'agitation thermique, il diffuse jusqu'a sa capture par un noyau. La separation de l'evolution d'un neutron en deux phases, phase de ralentissement, puis phase de diffusion, est, en realite, arbitraire; un neutron doit atteindre l'equilibre thermique progressivement. L'objet de ce rapport est de voir comment on peut deduire la repartition des thermiques de la repartition des rapides connaissant la repartition des thermiques naissants et la loi de diffusion. (au0010te.

  16. Nitrogen determination in wheat by neutron activation analysis using fast neutron flux from a thermal nuclear reactor

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

    This is a study of the technique for the determination of nitrogen and other elements in wheat flour through activation analysis with fast neutrons from a thermal nuclear reactor. The study begins with an introduction about the basis of the analytical methods, the equipment used in activation analysis and a brief description of the neutrons source. In the study are included the experiments carried out in order to determine the flux form in the site of irradiation, the N-13 half life and the interference due to the sample composition. (author)

  17. Polarized neutron reflectivity study of a thermally treated MnIr/CoFe exchange bias system.

    Science.gov (United States)

    Awaji, Naoki; Miyajima, Toyoo; Doi, Shuuichi; Nomura, Kenji

    2010-12-01

    It has recently been found that the exchange bias of a MnIr/CoFe system can be increased significantly by adding a thermal treatment to the bilayer. To reveal the origin of the higher exchange bias, we performed polarized neutron reflectivity measurements at the JRR-3 neutron source. The magnetization vector near the MnIr/CoFe interface for thermally treated samples differed from that for samples without the treatment. We propose a model in which the pinned spin area at the interface is extended due to the increased roughness and atomic interdiffusion that result from the thermal treatment.

  18. Neutronics and thermal-hydraulics analysis of KUHFR

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States); Mishima, K [KURRI, Osaka (Japan)

    1983-08-01

    control rod worth with reduced enrichment has not yet determined, but only a small decrease in worth is expected. These BOL boron poisoned fuels are also used as the fresh fuel feed for the equilibrium fuel cycle studies contained in this report. The first three cases shown have matching cycle lengths in the equilibrium cycle, while the last case has a considerably longer cycle length. These results are similarly reflected in the 'Maximum Cycle Lengths' shown for unpoisoned BOL cores. Thus, the first three case can be considered comparable. The last case might be considered as an option for an extended cycle length design. The cycle length for this case is increased by about 21%. Obviously, by decreasing the uranium density in the fuel meat (to 2.7 g/cm{sup 3}), the cycle length for this design could be reduced to match that of the other cases. Thermal-hydraulic calculations have been carried out in order to study the safety aspects of the use of reduced enrichment uranium fuel for the KUHFR. The calculations were based on what is outlined in the Safety Analysis Report for the KUHFR and also the IAEA Guidebook for the RERTR program. Only a few combinations of hydraulic parameters have been tested because the reactor safety cannot be discussed without any nuclear physics considerations. For example, any variations in fuel coolant channels may change not only flow velocities but also power peaking factors which may affect the assessment of reactor safety. For this reason, the thermal-hydraulic calculations were carried out only for those specific cases on which neutronics analysis has been already performed. Low enriched uranium (LEU) cases are also included in this study as initial feasibility studies for potential conversion. The computer code PLTEMP has been developed to calculate the flow distribution in the core, fuel plate temperatures and DNB heat fluxes.

  19. Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

    CERN Document Server

    Al-Jarallah, M I; Fazal-Ur-Rehman; Abu-Jarad, F A

    2002-01-01

    Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has bee...

  20. RBE of thermal neutrons for induction of chromosome aberrations in human lymphocytes.

    Science.gov (United States)

    Schmid, E; Wagner, F M; Canella, L; Romm, H; Schmid, T E

    2013-03-01

    The induction of chromosome aberrations in human lymphocytes irradiated in vitro with slow neutrons was examined to assess the maximum low-dose RBE (RBE(M)) relative to (60)Co γ-rays. For the blood irradiations, cold neutron beam available at the prompt gamma activation analysis facility at the Munich research reactor FRM II was used. The given flux of cold neutrons can be converted into a thermally equivalent one. Since blood was taken from the same donor whose blood had been used for previous irradiation experiments using widely varying neutron energies, the greatest possible accuracy was available for such an estimation of the RBE(M) avoiding the inter-individual variations or differences in methodology usually associated with inter-laboratory comparisons. The magnitude of the coefficient α of the linear dose-response relationship (α = 0.400 ± 0.018 Gy(-1)) and the derived RBE(M) of 36.4 ± 13.3 obtained for the production of dicentrics by thermal neutrons confirm our earlier observations of a strong decrease in α and RBE(M) with decreasing neutron energy lower than 0.385 MeV (RBE(M) = 94.4 ± 38.9). The magnitude of the presently estimated RBE(M) of thermal neutrons is-with some restrictions-not significantly different to previously reported RBE(M) values of two laboratories.

  1. Study of the Li2CO3 as thermal neutrons detector

    International Nuclear Information System (INIS)

    Herrera A, E.; Urena N, F.; Delfin L, A.

    2003-01-01

    The use every day but it frequents of the thermal neutrons in the treatment of tumours, using the neutron capture therapy technique in boron, there is generated the necessity to develop a dosimetric system that allows to evaluate in a reliable way the fluence and consequently the dose of neutrons that it is given in the tumours of the patients. One of the techniques but employees to determine the neutron fluence sub cadmic and epi cadmic in an indirect way, it is the activation of thin sheets of gold undress and covered with cadmium respectively that when being exposed to a neutron beam to the nuclear reaction 197 Au (n, γ ) 198 Au, emitting gamma radiation with an energy of 0.4118 MeV, being this, a disadvantage to be used as dosemeter. On the other hand, when exposing the lithium carbonate to a thermal neutron beam, free radicals of CO 3 that are quantified by the electron paramagnetic resonance technique are generated. This work analyzes those basic parameters that determine if those made up of Li 2 CO 3 complete with the requirements to be used as detectors and/or dosemeters of thermal neutrons. (Author)

  2. Gadolinium oxide coated fully depleted silicon-on-insulator transistors for thermal neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vitale, Steven A., E-mail: steven.vitale@ll.mit.edu; Gouker, Pascale M.

    2013-09-01

    Fully depleted silicon-on-insulator transistors coated with gadolinium oxide are shown to be effective thermal neutron dosimeters. The theoretical neutron detection efficiency is calculated to be higher for Gd{sub 2}O{sub 3} than for other practical converter materials. Proof-of-concept dosimeter devices were fabricated and tested during thermal neutron irradiation. The transistor current changes linearly with neutron dose, consistent with increasing positive charge in the SOI buried oxide layer generated by ionization from high energy {sup 157}Gd(n,γ){sup 158}Gd conversion electrons. The measured neutron sensitivity is approximately 1/6 the maximum theoretical value, possibly due to electron–hole recombination or conversion electron loss in interconnect wiring above the transistors. -- Highlights: • A novel Gd{sub 2}O{sub 3} coated FDSOI MOSFET thermal neutron dosimeter is presented. • Dosimeter can detect charges generated from {sup 157}Gd(n,γ){sup 158}Gd conversion electrons. • Measured neutron sensitivity is comparable to that calculated theoretically. • Dosimeter requires zero power during operation, enabling new application areas.

  3. Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

    Energy Technology Data Exchange (ETDEWEB)

    Al-Jarallah, M.I.; Naqvi, A.A. E-mail: aanaqvi@kfupm.edu.sa; Fazal-ur-Rehman; Abu-jarad, F

    2002-10-01

    Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has been found between the experimental results and the calculations.

  4. Filtered thermal neutron captured cross-sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Son, Pham Ngoc; Tan, Vuong Huu

    2014-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

  5. Absence of storage effects on radiation damage after thermal neutron irradiation of dry rice seeds

    Energy Technology Data Exchange (ETDEWEB)

    Kowyama, Y. [Mie Univ., Tsu (Japan); Saito, M.; Kawase, T.

    1987-09-15

    Storage effects on dry rice seeds equilibrated to 6.8% moisture content were examined after irradiation with X-rays of 5, 10, 20 and 40 kR and with thermal neutrons of 2.1, 4.2, 6.3 and 8.4×10{sup 13}N{sub th}/cm{sup 2}. Reduction in root growth was estimated from dose response curves after storage periods of 1 hr to 21 days. The longer the storage period, the greater enhancement of radiation damages in X-irradiated seeds. There were two components in the storage effect, i. e., a rapid increase of radiosensitivity within the first 24 hr and a slow increase up to 21 days. An almost complete absence of a storage effect was observed after thermal neutron exposure, in spite of considerably high radioactivities of the induced nuclides, {sup 56}Mn, {sup 42}K and {sup 24}Na, which were detected from gamma-ray spectrometry of the irradiated seeds. The present results suggest that the contributions of gamma-rays from the activated nuclides and of inherent contaminating gamma-rays are little or negligible against the neutron-induced damage, and that the main radiobiological effects of thermal neutrons are ascribed to in situ radiations, i, e., heavy particles resulting from neutron-capture reaction of atom. A mechanism underlying the absence of storage effect after thermal neutron irradiation was briefly discussed on the basis of radical formation and decay. (author)

  6. Analysis of cavity effect on space- and time-dependent fast and thermal neutron energy spectra

    International Nuclear Information System (INIS)

    Kudo, Katsuhisa; Narita, Masakuni; Ozawa, Yasutomo.

    1975-01-01

    The effects of the presence of a central cavity on the space- and time-dependent neutron energy spectra in both thermal and fast neutron systems are analyzed theoretically with use made of the multi-group one-dimensional time-dependent Ssub(n) method. The thermal neutron field is also analyzed for the case of a fundamental time eigenvalue problem with the time-dependent P 1 approximation. The cavity radius is variable, and the system radius for graphite is 120 cm and for the other materials 7 cm. From the analysis of the time-dependent Ssub(n) calculations in the non-multiplying systems of polythene, light water and graphite, cavity heating is the dominant effect for the slowing-down spectrum in the initial period following fast neutron burst, and when the slowing-down spectrum comes into the thermal energy region, cavity heating shifts to cavity cooling. In the multiplying system of 235 U, cavity cooling also takes place as the spectrum approaches equilibrium after the fast neutron burst is injected. The mechanism of cavity cooling is explained analytically for the case of thermal neutron field to illustrate its physical aspects, using the time-dependent P 1 approximation. An example is given for the case of light water. (auth.)

  7. Design of thermal neutron beam based on an electron linear accelerator for BNCT.

    Science.gov (United States)

    Zolfaghari, Mona; Sedaghatizadeh, Mahmood

    2016-12-01

    An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

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

    International Nuclear Information System (INIS)

    Mezei, F.

    1980-01-01

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

  9. The neutron silicon lens. An update of the thermal neutron lens results

    International Nuclear Information System (INIS)

    Johnson, M.W.; Daymond, M.R.

    2001-01-01

    This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

  10. The neutron silicon lens. An update of the thermal neutron lens results

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, M.W.; Daymond, M.R. [Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire (United Kingdom)

    2001-03-01

    This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

  11. Black holes and neutron stars: evolution of binary systems

    International Nuclear Information System (INIS)

    Kraft, R.P.

    1975-01-01

    Evidence for the existence of neutron stars and black holes in binary systems has been reviewed, and the following summarizes the current situation: (1) No statistically significant case has been made for the proposition that black holes and/or neutron stars contribute to the population of unseen companions of ordinary spectroscopic binaries; (2) Plausible evolutionary scenarios can be advanced that place compact X-ray sources into context as descendants of several common types of mass-exchange binaries. The collapse object may be a black hole, a neutron star, or a white dwarf, depending mostly on the mass of the original primary; (3) The rotating neutron star model for the pulsating X-ray sources Her X-1 and Cen X-3 is the simplest interpretation of these objects, but the idea that the pulsations result from the non-radial oscillations of a white dwarf cannot be altogether dismissed. The latter is particularly attractive in the case of Her X-1 because the total mass of the system is small; (4) The black hole picture for Cyg X-1 represents the simplest model that can presently be put forward to explain the observations. This does not insure its correctness, however. The picture depends on a long chain of inferences, some of which are by no means unassailable. (Auth.)

  12. Measurement of plutonium oxalate in thermal neutron coincidence counters

    International Nuclear Information System (INIS)

    Marshall, R.S.; Erkkila, B.H.

    1979-01-01

    A coincidence neutron counting method has been developed for assaying batches of plutonium oxalate. Using counting data from two concentric rings of 3 He detectors, corrections are made for the effects that water has on the coincidence neutron count rate. Batches of plutonium oxalate varying from 750 to 1000 g of plutonium and from 34 to 54% water are assayed with an average accuracy of +-3%

  13. Measurement of diffusion length of thermal neutrons in concrete

    International Nuclear Information System (INIS)

    Moser, M.

    2007-04-01

    The diffusion length of neutrons with a medium energy < 0.025 eV in concrete were determined using 4π-β detector and gamma detectors. Then it was possible to determine how deep can neutrons penetrate diverse concrete construction parts in a reactor in operation, with this method the dismantling process of a reactor can be planned in terms of what parts can be removed without danger and what parts can be assumed still are activated. (nevyjel)

  14. Contribution to solving the problem of neutron thermalization in heterogeneous reactor; Prilog resavanju problema termalizacije neutron u heterognom reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1963-12-15

    A method for calculating of neutron termalization in heterogeneous rector core was developed. It is more precise than the diffusion method but more complcated. Concerning accuracy it is comparable to non-diffusion methods. Sonce the approach was analytical need for powerful computer is avoided and the description of physical phenomena is more transparent. Convergence is satsfactory. Constraints of the proposed method are: low neutron absorption in the moderator, negligible slowing down in the fuel, and big lattice pitch. The method is applicable for heavy water and graphite moderator systems. Based on the application of this method, procedures were developed for calculating thermal utilzation and neutron temperature. Since 1/v dependence of cross sections is not estimated this metof could be used for long-term reactivity changes.

  15. Effects of Neutron Emission on Fragment Mass and Kinetic Energy Distribution from Thermal Neutron-Induced Fission of 235U

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Saetone, E.

    2007-01-01

    The mass and kinetic energy distribution of nuclear fragments from thermal neutron-induced fission of 235 U(n th ,f) have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening in the standard deviation of the kinetic energy at the final fragment mass number around m = 109, our simulation also produces a second broadening around m = 125. These results are in good agreement with the experimental data obtained by Belhafaf et al. and other results on yield of mass. We conclude that the obtained results are a consequence of the characteristics of the neutron emission, the sharp variation in the primary fragment kinetic energy and mass yield curves. We show that because neutron emission is hazardous to make any conclusion on primary quantities distribution of fragments from experimental results on final quantities distributions

  16. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2009-01-01

    A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

  17. CaSO{sub 4}:Dy microphosphor for thermal neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Bhadane, Mahesh S. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Mandlik, Nandkumar [Department of Physics, Fergusson College, Savitribai Phule Pune University, Pune 411007 (India); Patil, B.J. [Department of Physics, Abasaheb Garware College, Pune 411004 (India); Dahiwale, S.S.; Sature, K.R.; Bhoraskar, V.N. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India)

    2016-02-15

    Dysprosium-doped calcium sulphate (CaSO{sub 4}:Dy) microphosphor was synthesized by acid re-crystallization method and its thermoluminescence (TL) properties irradiated with thermal neutrons was studied. Structural and morphological characteristics have been studied using X-ray diffraction and SEM which mainly exhibits a orthorhombic structure with particle size of 200 to 250 µm. Moreover, thermal neutron dosimetric characteristics of the microphosphor such as thermoluminescence glow curve, TL dose–response have been studied. This microphosphor powder represents a TL glow peak (T{sub max}) centered at around 240 °C. The TL response of CaSO{sub 4}:Dy microphosphor as a function of thermal neutron fluence is observed to be very linear upto the fluence of 52×10{sup 11} n/cm{sup 2} and further saturates. In addition, TL glow curves were deconvoluted by computerized glow curve deconvolution (CGCD) method and corresponding trapping parameters have been determined. It has been found that for every deconvoluted peak there is change in the order of kinetics. Overall, the experimental results show that the CaSO{sub 4}:Dy microphosphor can have potential to be an effective thermal neutron dosimetry. - Highlights: • Acid-recrystallization method is used to prepare CaSO{sub 4}:Dy microphosphor • CaSO{sub 4}:Dy phosphor irradiated thermal neutrons for dosimetric application. • TL response curve showed to be a perfect linear. • Trapping parameters has been calculated using CGCD curve fitting.

  18. Sintering and microstructure evolution in columnar thermal barrier coatings

    International Nuclear Information System (INIS)

    Krishnamurthy, Ramanathan; Srolovitz, David J.

    2009-01-01

    Sintering of thermal barrier coatings changes their key properties, such as thermal conductivity and thermal shock resistance, thus adversely impacting their reliability. We present a novel modeling approach to study the evolution of coating structure during sintering. We model the sintering of individual columns using a thermodynamic principle, and incorporate the center-to-center approach rates for the columns calculated using this principle in a larger scale discrete dynamics model for the evolution of a large number of columns. Surface energies, grain boundary energies and strain energies associated with the deformation of the columns are all included in this framework, while sintering is assumed to occur by the concerted action of surface and grain boundary diffusion. Two sets of initial conditions corresponding to different extents of pre-sintering among neighboring columns are considered. When the extent of pre-sintering is small, we observe that small clusters containing 5-20 columns are formed. In contrast, where a larger amount of pre-sintering exists, we observe, especially at large column densities, that clusters containing 50-100 columns separated by large inter-cluster pores/channels that appear to organize themselves into a network are formed. These observations are in good agreement with recently published experimental observations. We also explain how these results can explain the development of a 'mud-crack'-like pattern

  19. Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

    International Nuclear Information System (INIS)

    Zhou, T.; Hawari, A. I.; Wehring, B. W.

    2007-01-01

    Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

  20. Simulation and optimisation of a position sensitive scintillation detector with wavelength shifting fibers for thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Herzkamp, Matthias; Engels, Ralf; Kemmerling, Guenter [ZEA-2, Forschungszentrum Juelich (Germany); Brueckel, Thomas [JCNS, Forschungszentrum Juelich (Germany); Stahl, Achim [III. Physikalisches Institut B, RWTH Aachen (Germany); Waasen, Stefan van [ZEA-2, Forschungszentrum Juelich (Germany); Faculty of Engineering, University of Duisburg-Essen (Germany)

    2015-07-01

    In neutron scattering experiments it is important to have position sensitive large scale detectors for thermal neutrons. A detector based on a neutron scintillator with wave length shifting fibers is a new kind of such a detector. We present the simulation of the detector based on the microscopic structure of the scintillation material of the mentioned detector. It consists of a converter and a scintillation powder bound in a matrix. The converter in our case is lithium fluoride with enriched lithium 6, to convert thermal neutrons into high energetic alpha and triton particles. The scintillation material is silver doped zinc sulfide. We show that pulse height spectra obtained by these scintillators can be be explained by the simple model of randomly distributed spheres of zinc sulfide and lithium fluoride. With this model, it is possible to optimise the mass ratio of zinc sulfide to lithium fluoride with respect to detection efficiency and/or energy deposition in zinc sulfide.

  1. Numerical Simulations of Pillar Structured Solid State Thermal Neutron Detector Efficiency and Gamma Discrimination

    Energy Technology Data Exchange (ETDEWEB)

    Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R

    2008-06-24

    This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  3. Electret ionization chamber: a new method for detection and dosimetry of thermal neutrons

    International Nuclear Information System (INIS)

    Ghilardi, A.J.P.

    1988-01-01

    An electret ionization chamber with boron coated walls is presented as a new method for detecting thermal neutrons. The efficiency of electret ionization chambers with different wall materials for the external electrode was inferred from the results. Detection of slow neutrons with discrimination against the detection of γ-rays and energetic neutrons was shown to depend on the selection of these materials. The charge stability over a long period of time and the charge decay owing to natural radiation were also studied. Numerical analysis was developed by the use of a micro-computer PC-XT. Both the experimental and numerical results show that the sensitivity of the electret ionization chamber for detection of thermal neutrons is comparable with that of the BF 3 ionization chamber and that new technologies for deposition of the boron layer will produce higher efficiency detectors. (author). 102 refs, 32 fig, 10 tabs

  4. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K.; Gabanska, B.; Igielski, A.; Krynicka, E.; Woznicka, U. [Institute of Nuclear Physics, Cracow (Poland)

    1993-12-31

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a {sup 3}He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs.

  5. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Gabanska, B.; Igielski, A.; Krynicka, E.; Woznicka, U.

    1993-01-01

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a 3 He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs

  6. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K; Gabanska, B; Igielski, A; Krynicka, E; Woznicka, U [Institute of Nuclear Physics, Cracow (Poland)

    1994-12-31

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a {sup 3}He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs.

  7. Enhancement of thermal neutron shielding of cement mortar by using borosilicate glass powder.

    Science.gov (United States)

    Jang, Bo-Kil; Lee, Jun-Cheol; Kim, Ji-Hyun; Chung, Chul-Woo

    2017-05-01

    Concrete has been used as a traditional biological shielding material. High hydrogen content in concrete also effectively attenuates high-energy fast neutrons. However, concrete does not have strong protection against thermal neutrons because of the lack of boron compound. In this research, boron was added in the form of borosilicate glass powder to increase the neutron shielding property of cement mortar. Borosilicate glass powder was chosen in order to have beneficial pozzolanic activity and to avoid deleterious expansion caused by an alkali-silica reaction. According to the experimental results, borosilicate glass powder with an average particle size of 13µm showed pozzolanic activity. The replacement of borosilicate glass powder with cement caused a slight increase in the 28-day compressive strength. However, the incorporation of borosilicate glass powder resulted in higher thermal neutron shielding capability. Thus, borosilicate glass powder can be used as a good mineral additive for various radiation shielding purposes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column

    International Nuclear Information System (INIS)

    Dou Haifeng; Dai Junlong

    2006-01-01

    In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)

  9. A large 2D PSD for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Knott, R.B.; Watt, G.; Boldeman, J.W. [Australian Nucl. Sci. and Tech. Organ., Menai, NSW (Australia). Phys. Div.; Smith, G.C. [Instrumentation Division, Brookhaven National Laboratory, Upton, NY 11973 (United States)

    1997-06-21

    A 2D PSD based on a MWPC has been constructed for a small angle neutron scattering instrument. The active area of the detector was 640 x 640 mm{sup 2}. To meet the specifications for neutron detection efficiency and spatial resolution, and to minimise parallax, the gas mixture was 190 kPa {sup 3}He plus 100 kPa CF{sub 4}, and the active volume had a thickness of 30 mm. The design maximum neutron count rate of the detector was 10{sup 5} events per second. The (calculated) neutron detection efficiency was 60% for 2 A neutrons and the (measured) neutron energy resolution on the anode grid was typically 20% (fwhm). The location of a neutron detection event within the active area was determined using the wire-by-wire method: the spatial resolution (5 x 5 mm{sup 2}) was thereby defined by the wire geometry. A 16-channel charge-sensitive preamplifier/amplifier/comparator module has been developed with a channel sensitivity of 0.1 V/fC, noise line width of 0.4 fC (fwhm) and channel-to-channel cross-talk of less than 5%. The proportional counter operating system (PCOS III) (LeCroy Corp, USA) was used for event encoding. The ECL signals produced by the 16 channel modules were latched in PCOS III by a trigger pulse from the anode and the fast encoders produce a position and width for each event. The information was transferred to a UNIX workstation for accumulation and online display. (orig.).

  10. A large 2D PSD for thermal neutron detection

    International Nuclear Information System (INIS)

    Knott, R.B.; Watt, G.; Boldeman, J.W.; Smith, G.C.

    1996-01-01

    A 2D PSD based on a MWPC has been constructed for a small angle neutron scattering instrument. The active area of the detector was 640 x 640 mm 2 . To meet the specifications for neutron detection efficiency and spatial resolution, and to minimize parallax, the gas mixture was 190 kPa 3 He plus 100 kPa CF 4 and the active volume had a thickness of 30 mm. The design maximum neutron count-rate of the detector was 10 5 events per second. The (calculated) neutron detection efficiency was 60% for 2 angstrom neutrons and the (measured) neutron energy resolution on the anode grid was typically 20% (fwhm). The location of a neutron detection event within the active area was determined using the wire-by-wire method: the spatial resolution (5 x 5 mm 2 ) was thereby defined by the wire geometry. A 16 channel charge-sensitive preamplifier/amplifier/comparator module has been developed with a channel sensitivity of 0.1 V/fC, noise linewidth of 0.4 fC (fwhm) and channel-to-channel cross-talk of less than 5%. The Proportional Counter Operating System (PCOS III) (LeCroy Corp USA) was used for event encoding. The ECL signals produced by the 16 channel modules were latched in PCOS III by a trigger pulse from the anode and the fast encoders produce a position and width for each event. The information was transferred to a UNIX workstation for accumulation and online display

  11. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.

    2015-01-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  12. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  13. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  14. Radioactive nuclides formed by irradiation of the natural elements with thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Ekberg, Kim

    1959-05-15

    For each natural element up to Bi this report gives: the 2200 m/sec neutron absorption cross section; the nuclides formed by thermal neutron activation; the saturation activity per gram natural element for a certain flux; half life and 'tenth life' of the activity; {beta}-energy and/or type of decay; mean {gamma} energy per disintegration; energy and abundance of {gamma} quanta.

  15. Thermal and fast neutron dosimetry using artificial single crystal diamond detectors

    International Nuclear Information System (INIS)

    Angelone, M.; Pillon, M.; Prestopino, G.; Marinelli, Marco; Milani, E.; Verona, C.; Verona-Rinati, G.; Aielli, G.; Cardarelli, R.; Santonico, R.; Bedogni, R.; Esposito, A.

    2011-01-01

    In this work we propose the artificial Single Crystal Diamond (SCD) detector covered with a thin layer (0.5 μm/4 μm) of 6 LiF as a simultaneous thermal and fast neutron fluence monitor. Some interesting properties of the diamond response versus the neutron energy are evidenced thanks to Monte Carlo simulation using the MCNPX code which allows to propose the diamond detector also as an ambient dose equivalent (H∗(10)) monitor (REM counter).

  16. Thermal neutron spectra measurements in IEAR-1 Reactor, by using a crystal spectrometer

    International Nuclear Information System (INIS)

    Fulfaro, R.; Figueiredo Neto, A.M.; Stasiulevicius, E.; Vinhas, L.A.

    1975-01-01

    The thermal neutron spectrum of the IEN Argonauta reactor has been measured in the wavelength from 0.7 to 1.9A, using a neutron crystal spectrometer. An aluminium single crystal, in transmission, was used as monochromator. The aluminium crystal reflectivity employed in the analysis of the data was calculated for the first five permitted orders. An effective absorption coefficient of the crystal was used to perform the calculations instead of the macroscopic cross section of the element

  17. Radioactive nuclides formed by irradiation of the natural elements with thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Ekberg, Kim

    1959-05-15

    For each natural element up to Bi this report gives: the 2200 m/sec neutron absorption cross section; the nuclides formed by thermal neutron activation; the saturation activity per gram natural element for a certain flux; half life and 'tenth life' of the activity; {beta}-energy and/or type of decay; mean {gamma} energy per disintegration; energy and abundance of {gamma} quanta.

  18. Solid thermoluminescent dosemeter of sodium tetraborate and brazilian fluoride sensitive to thermal neutrons

    International Nuclear Information System (INIS)

    Fratin, L.

    1988-01-01

    The techniques of compacting sodium tetraborate and natural fluoride mixtures were studied in this work, with the aim of producing a solid dosimeter sensitive to thermal neutrons. The production procedure involves the vitrification of the sodium tetraborate, the grinding, mixture, cold pressing and the sinterization of the pellets. A special arrangement was built for irradiation where paraffin was used as moderator for neutrons from a 241 Am-Be source. Two different mass ratios of sodium tetraborate and flourite showed a linear thermoluminescent response to the neutron fluence in the range of 1.0 to 7.0 x 10 8 n (sub)tcm -2 . Solid dosimeters, manufactured from natural fluorite and sodium chloride, showed a response to gamma radiation similar to the response of the dosimeters sensitive to neutrons. These dosimeters are need to identify the proportion of thermoluminescent response due to gamma radiation present in a neutron field. (author) [pt

  19. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

  20. Physical model of evolution of oxygen subsystem of PLZT-ceramics at neutron irradiation and annealing

    CERN Document Server

    Kulikov, D V; Trushin, Y V; Veber, K V; Khumer, K; Bitner, R; Shternberg, A R

    2001-01-01

    The physical model of evolution of the oxygen subsystem defects of the ferroelectric PLZT-ceramics by the neutron irradiation and isochrone annealing is proposed. The model accounts for the effect the lanthanum content on the material properties. The changes in the oxygen vacancies concentration, calculated by the proposed model, agree well with the polarization experimental behavior by the irradiated material annealing

  1. The Continued Spectral Evolution of the Neutron Star RX J0720.4-3125

    NARCIS (Netherlands)

    Vink, Jacco; de Vries, Cor P.; Méndez, Mariano; Verbunt, Frank

    2004-01-01

    We observed the isolated neutron star RX J0720.4-3125 with Chandra's Low Energy Transmission Grating Spectrometer, following the XMM-Newton discovery of the long-term spectral evolution of this source. The new observation shows that the spectrum of RX J0720.4-3125 has continued to change in the

  2. Evolution of the Distribution of Neutron Exposures in the Galaxy Disc ...

    Indian Academy of Sciences (India)

    this work was the development of an analytical model for the NEG, to provide ... paper is organized as follows: in section 2 we deduce the evolution equation of .... of neutron exposures for the seed nuclei that can experience any irradiation times. ..... We thank the referee for an extensive and helpful review, containing very ...

  3. Neutronics Study on LEU Nuclear Thermal Rocket Fuel Options

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  4. The thermal evolution of Mercury's Fe-Si core

    Science.gov (United States)

    Knibbe, Jurriën Sebastiaan; van Westrenen, Wim

    2018-01-01

    We have studied the thermal and magnetic field evolution of planet Mercury with a core of Fe-Si alloy to assess whether an Fe-Si core matches its present-day partially molten state, Mercury's magnetic field strength, and the observed ancient crustal magnetization. The main advantages of an Fe-Si core, opposed to a previously assumed Fe-S core, are that a Si-bearing core is consistent with the highly reduced nature of Mercury and that no compositional convection is generated upon core solidification, in agreement with magnetic field indications of a stable layer at the top of Mercury's core. This study also present the first implementation of a conductive temperature profile in the core where heat fluxes are sub-adiabatic in a global thermal evolution model. We show that heat migrates from the deep core to the outer part of the core as soon as heat fluxes at the outer core become sub-adiabatic. As a result, the deep core cools throughout Mercury's evolution independent of the temperature evolution at the core-mantle boundary, causing an early start of inner core solidification and magnetic field generation. The conductive layer at the outer core suppresses the rate of core growth after temperature differences between the deep and shallow core are relaxed, such that a magnetic field can be generated until the present. Also, the outer core and mantle operate at higher temperatures than previously thought, which prolongs mantle melting and mantle convection. The results indicate that S is not a necessary ingredient of Mercury's core, bringing bulk compositional models of Mercury more in line with reduced meteorite analogues.

  5. Water and the thermal evolution of carbonaceous chondrite parent bodies

    International Nuclear Information System (INIS)

    Grimm, R.E.; Mcsween, H.Y. Jr.

    1989-01-01

    Two hypotheses are proposed for the aqueous alteration of carbonaceous chondrites within their parent bodies, in which respectively the alteration occurs (1) throughout the parent body interior, or (2) in a postaccretional surface regolith; both models assume an initially homogeneous mixture of ice and rock that is heated through the decay of Al-26. Water is seen to exert a powerful influence on chondrite evolution through its role of thermal buffer, permitting substitution of a low temperature aqueous alteration for high temperature recrystallization. It is quantitatively demonstrated that liquid water may be introduced by either hydrothermal circulation, vapor diffusion from below, or venting due to fracture. 104 refs

  6. Thermal evolution of the Schwinger model with matrix product operators

    International Nuclear Information System (INIS)

    Banuls, M.C.; Cirac, J.I.; Cichy, K.; Jansen, K.; Saito, H.

    2015-10-01

    We demonstrate the suitability of tensor network techniques for describing the thermal evolution of lattice gauge theories. As a benchmark case, we have studied the temperature dependence of the chiral condensate in the Schwinger model, using matrix product operators to approximate the thermal equilibrium states for finite system sizes with non-zero lattice spacings. We show how these techniques allow for reliable extrapolations in bond dimension, step width, system size and lattice spacing, and for a systematic estimation and control of all error sources involved in the calculation. The reached values of the lattice spacing are small enough to capture the most challenging region of high temperatures and the final results are consistent with the analytical prediction by Sachs and Wipf over a broad temperature range.

  7. Rifting and thermal evolution of the Northwestern Mediterranean

    Directory of Open Access Journals (Sweden)

    P. Chiozzi

    1995-06-01

    Full Text Available The structural setting of the Northwestern Mediterranean stems from tectonothermal processes which reflect on the nature of the crust. The Oligocene to Present evolution is here analysed with a thermal model which takes into account the significant extension of the continentallithosphere before the onset of sea-floor spread- ing in the bathyal zone. Subsidence data were used to set the boundaries of the oceanic realm which was com- pared with previous reconstructions inferred from other geophysical evidence. The thermal features of the transitional crust that lies between the oceanic crust and the stretched continental margins were also outlined. The Ligurian-Proven~al basin is a marginaI basin, whereas only the continental crust is expected in the Valen- cia trough. An evolutionary sketch of the study area that accounts for the observed subsidence and heat flux is proposed.

  8. Absolute efficiency calibration of 6LiF-based solid state thermal neutron detectors

    Science.gov (United States)

    Finocchiaro, Paolo; Cosentino, Luigi; Lo Meo, Sergio; Nolte, Ralf; Radeck, Desiree

    2018-03-01

    The demand for new thermal neutron detectors as an alternative to 3He tubes in research, industrial, safety and homeland security applications, is growing. These needs have triggered research and development activities about new generations of thermal neutron detectors, characterized by reasonable efficiency and gamma rejection comparable to 3He tubes. In this paper we show the state of the art of a promising low-cost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. A few configurations were studied with the GEANT4 simulation code, and the intrinsic efficiency of the corresponding detectors was calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured intrinsic detection efficiency is well reproduced by the simulations, therefore validating the simulation tool in view of new designs. These neutron detectors have also been tested at neutron beam facilities like ISIS (Rutherford Appleton Laboratory, UK) and n_TOF (CERN) where a few samples are already in operation for beam flux and 2D profile measurements. Forthcoming applications are foreseen for the online monitoring of spent nuclear fuel casks in interim storage sites.

  9. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Liu, H.B.

    1996-01-01

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 10 10 n/cm 2 ·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10 -11 cGy·cm 2 /n th ) and fast neutron (0.58 x 10 -11 cGy·cm 2 /n th ) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

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

    International Nuclear Information System (INIS)

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

    1995-08-01

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  11. Measured Thermal and Fast Neutron Fluence Rates for ATF-1 Holders During ATR Cycle 157D

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Larry Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, David Torbet [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 157D which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates. The fluence rates reported are for the average power levels given in the table of power history and distribution.

  12. Preparation of rock samples for measurement of the thermal neutron macroscopic absorption cross-section

    International Nuclear Information System (INIS)

    Czubek, J.A.; Burda, J.; Drozdowicz, K.; Igielski, A.; Kowalik, W.; Krynicka-Drozdowicz, E.; Woznicka, U.

    1986-03-01

    Preparation of rock samples for the measurement of the thermal neutron macroscopic absorption cross-section in small cylindrical two-region systems by a pulsed technique is presented. Requirements which should be fulfilled during the preparation of the samples due to physical assumptions of the method are given. A cylindrical vessel is filled with crushed rock and saturated with a medium strongly absorbing thermal neutrons. Water solutions of boric acid of well-known macroscopic absorption cross-section are used. Mass contributions of the components in the sample are specified. This is necessary for the calculation of the thermal neutron macroscopic absorption cross-section of the rock matrix. The conditions necessary for assuring the required accuracy of the measurement are given and the detailed procedure of preparation of the rock sample is described. (author)

  13. Fast neutron irradiation and thermal properties of doped nonstoichiometric lithium potassium sulphate crystals

    International Nuclear Information System (INIS)

    Kassem, M.E.; Gomaa, N.G.; El-Khatib, A.M.

    1990-01-01

    The influence of point defects introduced by fast neutron irradiations with neutron fluences up to 1.08 x 10 10 n/cm 2 on the thermal properties of pure and doped Li 1.4 K 0.6 SO 4 single crystals are studied in the vicinity of high temperature phase transition at 705 K. The temperature dependence of specific heat is found to be shifted towards lower temperature with the increase of neutron fluence, and can be affected by the presence of Cu 2+ dopant. The change in the value of the specific heat can be attributed to the presence of internal strain generated inside the crystal. (author)

  14. The dynamic method for time-of-flight measurement of thermal neutron spectra from pulsed sources

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Chuklyaev, S.V.; Tulaev, A.B.; Bobrakov, V.F.

    1995-01-01

    A time-of-flight method for measurement of thermal neutron spectra in pulsed neutron sources with an efficiency more than 10 5 times higher than the standard method is described. The main problems associated with the electric current technique for time-of-flight spectra measurement are examined. The methodical errors, problems of special neutron detector design and other questions are discussed. Some experimental results for spectra from the surfaces of water and solid methane moderators obtained at the IBR-2 pulsed reactor (Dubna, Russia) are presented. (orig.)

  15. Evolution of thermal fatigue management of piping in US LWRs

    International Nuclear Information System (INIS)

    McDewitt, M.; Wolfe, K.; McGill, R.

    2015-01-01

    Fatigue usage caused by cyclic changes of thermally stratified reactor coolant in Light Water Reactor (LWR) pressure boundary piping was not an original consideration in US Nuclear Power Plant (NPP) designs. During the mid 1980's, several events involving cracking and leakage due to thermal cycling occurred in reactor coolant system branch piping at both US and International NPPs. In 1988, the US Nuclear Regulatory Commission (US NRC) issued Bulletin 88-08 to alert LWR licensees of the potential for piping failures due to stratified thermal cycling. In response to these events, the US nuclear industry developed initiatives to identify susceptible components and established measures to monitor and prevent future failures. These initiatives have been effective in preventing leakage events, but have also identified fewer defects than expected based on screening model predictions. Improved analytical techniques are being investigated to maintain program effectiveness while minimizing unnecessary non-destructive examinations. This paper discusses the evolution of the US thermal fatigue initiatives, and analytical concepts being evaluated to improve program efficiency. (authors)

  16. Optimization of thermal neutron shield concrete mixture using artificial neural network

    Energy Technology Data Exchange (ETDEWEB)

    Yadollahi, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Nazemi, E., E-mail: nazemi.ehsan@yahoo.com [Young Researchers and Elite Club, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of); Zolfaghari, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Ajorloo, A.M. [Water and Environmental Engineering Department, Shahid Beheshti University, P.O. Box: 167651719, Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m{sup 3}, a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

  17. Optimization of thermal neutron shield concrete mixture using artificial neural network

    International Nuclear Information System (INIS)

    Yadollahi, A.; Nazemi, E.; Zolfaghari, A.; Ajorloo, A.M.

    2016-01-01

    Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m 3 , a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

  18. Effect of Different Thermal Neutron Fluxes on Blood of Male Mice

    International Nuclear Information System (INIS)

    Abd El-Latif, A.A.; Saeid, Kh. S.; Abd El-Latif, A.A.; Emara, N.M.; Emara, N.M.

    2010-01-01

    This work deals with the exposing of male mice to different fluxes of thermal neutron .Investigation has been performed by calculating of thermal neutron fluxes(0.27x10 8 N/cm 2 . 1h , 0.54x10 8 N/cm 2 . 1h, 1.08x10 8 N/cm 2 . 1h, 2.16x10 8 N/cm 2 . 3h and 4.32x10 8 N/cm 2 . 6h) which emitted from neutron irradiation cell with source Ra - Be (α,n) have activity 3 m. Ci made by leybold(55930) . The number and differential leucocytes counts types of white blood cells in million per cubic millimeter (W. B. Cs. mm -3 ) ,the number of platelets mm -3 ,the number of red blood cells in million per cubic millimeter (R. B. Cs. mm -3 ), the hemoglobin in Blood (mg/dl), the lymphocytes ,and the eosiniphil leucocytes in blood decrease with increasing thermal neutron fluxes. But neutrophile and monocytes in blood increase with increasing the thermal neutron fluxes

  19. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  20. Evaluation of thermal neutron cross-sections and resonance integrals of protactinium, americium, curium, and berkelium isotopes

    International Nuclear Information System (INIS)

    Belanova, T.S.

    1994-12-01

    Data on the thermal neutron fission and capture cross-sections as well as their corresponding resonance integrals are reviewed and analysed. The data are classified according to the form of neutron spectra under investigation. The weighted mean values of the cross-sections and resonance integrals for every type of neutron spectra were adopted as evaluated data. (author). 87 refs, 2 tabs

  1. Facility at CIRUS reactor for thermal neutron induced prompt γ-ray spectroscopic studies

    International Nuclear Information System (INIS)

    Biswas, D.C.; Danu, L.S.; Mukhopadhyay, S.; Kinage, L.A.; Prashanth, P.N.; Goswami, A.; Sahu, A.K.; Shaikh, A.M.; Chatterjee, A.; Choudhury, R.K.; Kailas, S.

    2013-01-01

    A facility for prompt γ-ray spectroscopic studies using thermal neutrons from a radial beam line of Canada India Research Utility Services (CIRUS) reactor, Bhabha Atomic Research Centre (BARC), has been developed. To carry out on-line spectroscopy experiments, two clover germanium detectors were used for the measurement of prompt γ rays. For the first time, the prompt γ–γ coincidence technique has been used to study the thermal neutron induced fission fragment spectroscopy (FFS) in 235 U(n th , f). Using this facility, experiments have also been carried out for on-line γ-ray spectroscopic studies in 113 Cd(n th , γ) reaction

  2. Light-water-reactor coupled neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1982-01-01

    An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

  3. Independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239Pu

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1975-01-01

    The relative independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239 Pu have been measured on line using a mass spectrograph and thermalized neutrons from a burst reactor. Independent yields were derived by normalizing the measurements to products of chain yields and fractional independent yields, estimating the latter from measured cumulative yields of Kr and Xe. Comparing the independent yields with those from 238 U fission, the 239 Pu results show shifts in isotopic yield distribution toward lower mass for both Rb and Cs and also toward the production of more Cs and less Rb when 239 Pu is fissioned

  4. Influence of the effective mass of water molecule on thermal neutron scattering

    International Nuclear Information System (INIS)

    Markovic, M.

    1981-01-01

    The influence of the effective water molecule mass on the thermal neutron scattering on the nucleus of the hydrogen atom has been investigated. Besides the actual water molecule mass (M = 18) the investigations have been carried out with its two effective values (M1 = 16 and M2 = 20). The differential and total cross sections have been calculated for the incident thermal neutron energy E o = 1 eV. Investigation results show different prominence of the quantum effects and for M2 the appearance of peaks in the quasielastic scattering. (author)

  5. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-02-15

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

  6. Measuring thermal neutron spectra of RIEN-1 reactor with a chopper

    International Nuclear Information System (INIS)

    Jesus Vilar, G. de.

    1977-03-01

    The setting up of a time-of-flight spectrometer (Fermi Chopper) and its use in measurements of thermal neutron spectra in the irradiation channels of the Argonaut Reactor(Instituto de Engenharia Nuclear, Brazil), is described. These distributions are obtained using a multichannel analyser with the necessary corrections being made for counting losses in the analyser, dectector efficiency experimental resolution and chopper transmission function. The results obtained show that the thermal neutron flux emerging from the canal J-9 can be approximately described by a Maxwellian distribution with and associated characteristic temperature fo 430+-30 0 K [pt

  7. Parity non-conservation in the capture of polarized thermal neutrons

    DEFF Research Database (Denmark)

    Warming, Inge Elisabeth

    1969-01-01

    The asymmetry in the intensity of γ-radiation following the capture of polarized thermal neutrons in 113Cd has been measured with Ge(Li) detectors. The result, A = (−0.6±1.8)×10−4, like that previously reported [1], gives no evidence for a non-zero effect.......The asymmetry in the intensity of γ-radiation following the capture of polarized thermal neutrons in 113Cd has been measured with Ge(Li) detectors. The result, A = (−0.6±1.8)×10−4, like that previously reported [1], gives no evidence for a non-zero effect....

  8. The Cross-Section for the Radiative Capture of Thermal Neutrons by Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1942-07-01

    This report is based on an experiment performed at the Cavendish Laboratory (Cambridge) by E. Broda, J. Guéron and L. Kowarski in July 1942 where the intensity of the beta-activity induced in uranium by thermal neutrons has been compared with that induced in manganese or iodine. Care was taken to avoid losses due to a Szilard-Chalmers effect. The capture cross section of uranium for thermal neutrons is found to amount to (2.78 ±0.1)*10{sup -24} cm{sup 2}, assuming the value 581*10{sup -24} cm{sup 2} for σ{sub B}. (nowak)

  9. Microchemical evolution of neutron-irradiated stainless steel

    International Nuclear Information System (INIS)

    Brager, H.R.; Garner, F.A.

    1980-04-01

    The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations

  10. The optical potential for thermal neutrons in second approximation

    International Nuclear Information System (INIS)

    Nowak, E.

    1980-01-01

    After the construction of an expansion of the transfer operator for neutron scattering in terms of the scattering length the optical potential for coherent scattering is derived whereby the corrections due to the dynamic properties of the scattering system are discussed. (HSI) [de

  11. Neutron production and thermal moderation at the PSI UCN source

    Energy Technology Data Exchange (ETDEWEB)

    Becker, H. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Institute for Particle Physics, Eidgenössische Technische Hochschule, Zürich (Switzerland); Bison, G.; Blau, B.; Chowdhuri, Z.; Eikenberg, J.; Fertl, M. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kirch, K. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Institute for Particle Physics, Eidgenössische Technische Hochschule, Zürich (Switzerland); Lauss, B., E-mail: bernhard.lauss@psi.ch [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Perret, G.; Reggiani, D.; Ries, D.; Schmidt-Wellenburg, P. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Talanov, V., E-mail: vadim.talanov@psi.ch [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Wohlmuther, M.; Zsigmond, G. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

    2015-03-21

    We report on gold foil activation measurements performed along a vertical channel along the tank of the ultracold neutron source at the Paul Scherrer Institute. The activities obtained at various distances from the spallation target are in very good agreement with MCNPX simulations which take into account the detailed description of the source as built.

  12. The measurements of thermal neutron flux distribution in a paraffin ...

    Indian Academy of Sciences (India)

    outputs and the best agreement was generated using NaI(Tl) by a minimum discrepancy of .... the four possible reactions with neutron absorption, our desire is to form 116m1In, which .... According to Geiger user's manual, its efficiency for.

  13. Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

    International Nuclear Information System (INIS)

    Noorddin Ibrahim; Rosnie Akang

    2009-01-01

    Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

  14. Thermal neutron capture and resonance integral cross sections of {sup 45}Sc

    Energy Technology Data Exchange (ETDEWEB)

    Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Thi Hien, Nguyen [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Guinyun, E-mail: gnkim@knu.ac.kr [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Kwangsoo [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shin, Sung-Gyun; Cho, Moo-Hyun [Department of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Lee, Manwoo [Research Center, Dongnam Institute of Radiological and Medical Science, Busan 619-953 (Korea, Republic of)

    2015-11-01

    The thermal neutron cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been measured relative to that of the {sup 197}Au(n,γ){sup 198}Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (G{sub th}) and resonance (G{sub epi}) neutron self-shielding, the γ-ray attenuation (F{sub g}) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been determined relative to the reference values of the {sup 197}Au(n,γ){sup 198}Au reaction, with σ{sub o,Au} = 98.65 ± 0.09 barn and I{sub o,Au} = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σ{sub o,Sc} = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be I{sub o,Sc} = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

  15. Time evolution of damage in thermally induced creep rupture

    KAUST Repository

    Yoshioka, N.

    2012-01-01

    We investigate the time evolution of a bundle of fibers subject to a constant external load. Breaking events are initiated by thermally induced stress fluctuations followed by load redistribution which subsequently leads to an avalanche of breakings. We compare analytic results obtained in the mean-field limit to the computer simulations of localized load redistribution to reveal the effect of the range of interaction on the time evolution. Focusing on the waiting times between consecutive bursts we show that the time evolution has two distinct forms: at high load values the breaking process continuously accelerates towards macroscopic failure, however, for low loads and high enough temperatures the acceleration is preceded by a slow-down. Analyzing the structural entropy and the location of consecutive bursts we show that in the presence of stress concentration the early acceleration is the consequence of damage localization. The distribution of waiting times has a power law form with an exponent switching between 1 and 2 as the load and temperature are varied.

  16. Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

    2009-07-01

    The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  18. Monte Carlo simulation of a coded-aperture thermal neutron camera

    International Nuclear Information System (INIS)

    Dioszegi, I.; Salwen, C.; Forman, L.

    2011-01-01

    We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

  19. Thermal evolution of nitrate precursors for processing of lanthanide perovskites

    Directory of Open Access Journals (Sweden)

    Kozhukharov, V. S.

    1998-12-01

    Full Text Available Studies on thermal decomposition of ceramic powder with a general formula of (La1-x Ba x (Co0.8 Fe0.2O3 have been achieved. Precursors as nitrate solutions with additive of EDTA as complexion agent are used for powder processing. The black powders obtained are dried and their thermal evolution up to 1000ºC has been investigated by Differential Thermal Analysis. The powders was analyzed by EDX and ICP- AES, as well. It was established that the powder compositions are very close to the nominal one. The resulting DTA, TA, TG and DTG curves are analyzed as function of the composition and heating rate applied. At polythermal scanning regime three regions the powder thermal evolution are discussed. The correlation dependence has been examined for both Sr- and Ba- doped multicomponent lanthanide samples. The multicomponent nature of the samples have been shown on the base of the thermal treatment applied and XRD phase control carried out.

    Se han realizado estudios sobre la descomposición térmica de polvos cerámicos de fórmula general (La1-x Ba x (Co0.8 Fe0.2O3. Se utilizaron como precursores soluciones de nitratos con EDTA como agente acomplejante. La evolución térmica del polvo negro obtenido se estudió hasta la temperatura de 1000 ºC por medio de análisis térmico diferencial. Los polvos se analizaron así mismo por EDX e ICP-A ES. Se estableció que la composición de los polvos esta muy próxima a la composición nominal. Se distingue tres regímenes en la evolución térmica. Se examina la dependencia con el contenido en lantanidas multicomponentes de pulsos con Sr y Ba. La naturaleza multicomponente se ha mostrado sobre la base del tratamiento térmico empleado y el análisis de las fases cristalinas.

  20. Measurement of thermal, epithermal and fast neutrons fluxes by the activation foil method at IEA-R1 reactor

    International Nuclear Information System (INIS)

    Dias, M.S.; Koskinas, M.F.; Berretta, J.R.; Fratin, L.; Botelho, S.

    1990-01-01

    The thermal, epithermal and fast neutron fluxes have been determined experimentally by the activation foil method at position GI, located near the IEA-R1 reactor core. The reactions used were 197 Au (n,gamma) 198 Au, for thermal and epithermal neutrons and 27 Na (n,alpha) 24 Na, for fast neutrons. The activities were measured by the 4π(PC)β-γ coincidence method. (author)

  1. Temperature dependence of the thermal expansion of neutron-irradiated pyrolytic carbon and graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1988-01-01

    The effects of neutron irradiation and annealing on the temperature dependence of the linear thermal expansion of pyrolytic carbon and graphite were investigated after irradiation at 930-1280 0 C to a maximum neutron fluence of 2.84 x 10 25 m -2 (E > 29 fJ). After irradiation, little change in the thermal expansion of pyrolytic graphite was observed. However, as-deposited pyrolytic carbon showed an increase in thermal expansion in the perpendicular direction, a decrease in the direction parallel to the deposition plane, and also an increase in the anisotropy of the thermal expansion. Annealing at 2000 0 C did not cause any effective changes for irradiated specimens of either as-deposited pyrolytic carbon or pyrolytic graphite. (author)

  2. ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

    Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

  3. Influence of an SN solver in a fine-mesh neutronics/thermal-hydraulics framework

    International Nuclear Information System (INIS)

    Jareteg, Klas; Vinai, Paolo; Demaziere, Christophe; Sasic, Srdjan

    2015-01-01

    In this paper a study on the influence of a neutron discrete ordinates (S N ) solver within a fine-mesh neutronic/thermal-hydraulic methodology is presented. The methodology consists of coupling a neutronic solver with a single-phase fluid solver, and it is aimed at computing the two fields on a three-dimensional (3D) sub-pin level. The cross-sections needed for the neutron transport equations are pre-generated using a Monte Carlo approach. The coupling is resolved in an iterative manner with full convergence of both fields. A conservative transfer of the full 3D information is achieved, allowing for a proper coupling between the neutronic and the thermal-hydraulic meshes on the finest calculated scales. The discrete ordinates solver is benchmarked against a Monte Carlo reference solution for a two-dimensional (2D) system. The results confirm the need of a high number of ordinates, giving a satisfactory accuracy in k eff and scalar flux profile applying S 16 for 16 energy groups. The coupled framework is used to compare the S N implementation and a solver based on the neutron diffusion approximation for a full 3D system of a quarter of a symmetric, 7x7 array in an infinite lattice setup. In this case, the impact of the discrete ordinates solver shows to be significant for the coupled system, as demonstrated in the calculations of the temperature distributions. (author)

  4. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1986-09-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data. 8 refs

  5. 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....

  6. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; van der Gucht, C.; Vanuxeem, J.

    1986-01-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processes to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. The authors show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data

  7. Neutronics and thermal hydraulics coupling scheme for design improvement of liquid metal fast systems

    International Nuclear Information System (INIS)

    Sanchez-Espinoza, V.H.; Jaeger, W.; Travleev, A.; Monti, L.; Doern, R.

    2009-01-01

    Many advanced reactor concepts are nowadays under investigations within the Generation IV international initiative as well as in European research programs including subcritical and critical fast reactor systems cooled by liquid metal, gas and supercritical water. The Institute of Neutron Physics and Reactor Technology (INR) at the Forschungszentrum Karlsruhe GmbH is involved in different European projects like IP EUROTRANS, ELSY, ESFR. The main goal of these projects is, among others, to assess the technical feasibility of proposed concepts regarding safety, economics and transmutation requirements. In view of increased computer capabilities, improved computational schemes, where the neutronic and the thermal hydraulic solution is iteratively coupled, become practicable. The codes ERANOS2.1 and TRACE are being coupled to analyze fuel assembly or core designs of lead-cooled fast reactors (LFR). The neutronic solution obtained with the coupled system for a LFR fuel assembly was compared with the MCNP5 solution. It was shown that the coupled system is predicting physically sound results. The iterative coupling scheme was realized using Perlscripts and auxiliary Fortran programs to ensure that the mapping between the neutronic and the thermal hydraulic part is consistent. The coupled scheme is very flexible and appropriate for the neutron physical and thermal hydraulic investigation of fuel assemblies and of cores of lead cooled fast reactors. The developed methods and the obtained results will be presented and discussed. (author)

  8. Bis(pinacolato)diboron as an additive for the detection of thermal neutrons in plastic scintillators

    Energy Technology Data Exchange (ETDEWEB)

    Mahl, Adam [Department of Physics and the Nuclear Science and Engineering Center (NuSEC), Colorado School of Mines, Golden, CO 80401 (United States); Yemam, Henok A.; Stuntz, John [Department of Chemistry and Geochemistry and the Materials Science Program Colorado School of Mines, Golden, CO 80401 (United States); Remedes, Tyler [Department of Physics and the Nuclear Science and Engineering Center (NuSEC), Colorado School of Mines, Golden, CO 80401 (United States); Sellinger, Alan [Department of Chemistry and Geochemistry and the Materials Science Program Colorado School of Mines, Golden, CO 80401 (United States); Greife, Uwe, E-mail: ugreife@mines.edu [Department of Physics and the Nuclear Science and Engineering Center (NuSEC), Colorado School of Mines, Golden, CO 80401 (United States)

    2016-04-21

    A readily available and inexpensive boron compound was tested as an additive for the detection of thermal neutrons in plastic scintillators. Bis(pinacolato)diboron (B{sub 2}Pin{sub 2}) was determined to be a compatible boron source (8.51 wt% boron, 1.70 wt% {sup 10}B) in poly(vinyltoluene) based matrices. Plastic scintillator blends of 1–20 wt% 2,5-diphenyloxazole (PPO), 0.1 wt% 1,4-bis(5-phenyloxazol-2-yl) benzene (POPOP) and 1–15 wt% B{sub 2}Pin{sub 2} were prepared that provided optical clarity, good mechanical properties, and the capability of thermal neutron detection. Independent of B{sub 2}Pin{sub 2} concentration, strong {sup 10}B neutron capture signals around 90 keV{sub ee} were observed at essentially constant light output. Increasing PPO concentration allowed for the use of pulse shape discrimination (PSD) in both fast and thermal neutron detection. High PPO concentrations appear to cause additional alpha quenching that affected the {sup 10}B neutron capture signal. Aging effects after storage in air for several months were observed, which led to degradation of performance and in some samples of mechanical stability.

  9. MCFT: a program for calculating fast and thermal neutron multigroup constants

    International Nuclear Information System (INIS)

    Yang Shunhai; Sang Xinzeng

    1993-01-01

    MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

  10. A novel track density measurement method by thermal neutron activation of DYECETs

    International Nuclear Information System (INIS)

    Sohrabi, M.; Mahdi, Sh.

    1995-01-01

    A novel track density evaluation method based on thermal neutron activation of some elements of dyed electrochemically etched tracks (DYECETs) of charged particles in detectors like polycarbonate (PC) followed by measurements of gamma activity of the activated detectors is introduced. In this method, the tracks of charged particles like fast neutron-induced recoils in PC detectors were electrochemically etched, dyed by ''QuicDYECET'' methods as recently introduced by us, activated by thermal neutrons and counted for gamma activity determination to be correlated with track density. The activities of elements such as bromine-82 ( 82 Br) and sodium-24 ( 24 Na) on dyes such as Eosin Yellowish, Eosin Bluish, etc. determined by a hyper-pure germanium detector, were found to be in good correlation with DYECET density and thus particle fluence or dose. The effects of different types of dyes and their elements, dye concentration, neutron fluences and ECE durations on the DYECET density responses were studied. This new development is a method of scientific interest, potentially possessing some interesting features, as an alternative method for ECE track density determination using a gamma activity measuring system. It also has the drawback of being applicable only in centres having thermal neutron facilities. The results of the above studies are presented and discussed. (Author)

  11. Bis(pinacolato)diboron as an additive for the detection of thermal neutrons in plastic scintillators

    International Nuclear Information System (INIS)

    Mahl, Adam; Yemam, Henok A.; Stuntz, John; Remedes, Tyler; Sellinger, Alan; Greife, Uwe

    2016-01-01

    A readily available and inexpensive boron compound was tested as an additive for the detection of thermal neutrons in plastic scintillators. Bis(pinacolato)diboron (B_2Pin_2) was determined to be a compatible boron source (8.51 wt% boron, 1.70 wt% "1"0B) in poly(vinyltoluene) based matrices. Plastic scintillator blends of 1–20 wt% 2,5-diphenyloxazole (PPO), 0.1 wt% 1,4-bis(5-phenyloxazol-2-yl) benzene (POPOP) and 1–15 wt% B_2Pin_2 were prepared that provided optical clarity, good mechanical properties, and the capability of thermal neutron detection. Independent of B_2Pin_2 concentration, strong "1"0B neutron capture signals around 90 keV_e_e were observed at essentially constant light output. Increasing PPO concentration allowed for the use of pulse shape discrimination (PSD) in both fast and thermal neutron detection. High PPO concentrations appear to cause additional alpha quenching that affected the "1"0B neutron capture signal. Aging effects after storage in air for several months were observed, which led to degradation of performance and in some samples of mechanical stability.

  12. Thermalization time in a model of neutron star

    Czech Academy of Sciences Publication Activity Database

    Ducomet, B.; Nečasová, Šárka

    2011-01-01

    Roč. 16, č. 3 (2011), s. 801-818 ISSN 1531-3492 R&D Projects: GA ČR GA201/08/0012 Institutional research plan: CEZ:AV0Z10190503 Keywords : comressible * heat conducting fluids * one-dimensional symmetry * neutron star Subject RIV: BA - General Mathematics Impact factor: 0.921, year: 2011 http://aimsciences.org/journals/displayArticlesnew.jsp?paperID=6325

  13. Neutron spectrometry and dosimetry by means of evolutive neural networks

    International Nuclear Information System (INIS)

    Ortiz R, J.M.; Martinez B, M.R.; Vega C, H.R.

    2008-01-01

    The artificial neural networks and the genetic algorithms are two relatively new areas of research, which have been subject to a growing interest during the last years. Both models are inspired by the nature, however, the neural networks are interested in the learning of a single individual, which is defined as fenotypic learning, while the evolutionary algorithms are interested in the adaptation of a population to a changing environment, that which is defined as genotypic learning. Recently, the use of the technology of neural networks has been applied with success in the area of the nuclear sciences, mainly in the areas of neutron spectrometry and dosimetry. The structure (network topology), as well as the learning parameters of a neural network, are factors that contribute in a significant way with the acting of the same one, however, it has been observed that the investigators in this area, carry out the selection of the network parameters through the essay and error technique, that which produces neural networks of poor performance and low generalization capacity. From the revised sources, it has been observed that the use of the evolutionary algorithms, seen as search techniques, it has allowed him to be possible to evolve and to optimize different properties of the neural networks, just as the initialization of the synaptic weights, the network architecture or the training algorithms without the human intervention. The objective of the present work is focused in analyzing the intersection of the neural networks and the evolutionary algorithms, analyzing like it is that the same ones can be used to help in the design processes and training of a neural network, this is, in the good selection of the structural parameters and of network learning, improving its generalization capacity, in such way that the same one is able to reconstruct in an efficient way neutron spectra and to calculate equivalent doses starting from the counting rates of a Bonner sphere

  14. Detection of plastic explosives using thermal neutron radiography

    International Nuclear Information System (INIS)

    Hacidume, Leo Ryoske

    1999-12-01

    The work aims to demonstrate the potentiality of the neutron radiography technique, allied to the computerized tomography by transmission, to both detect and visualize plastic explosive samples in several hidden conditions, using a simple scanner as a digitalisation instrument. Each tomographic essay was obtained in the J-9 channel of the Argonauta Research Reactor of IEN/CNEN, in groups of six neutron radiographic projections, performed with an angular increment of 30 deg C, in a period of time of 30 minutes for each projection. Two groups of tomographic reconstructions were generated, distinguished by the digitalisation process of the interested lines in the reconstruction plane coming from the projection groups, utilization a scanner and a microdensitometer, respectively. The reconstruction of the bi-dimensional image of the transverse section, in relation to this plane, was processed making use of the Image Reconstruction Algorithmic of an Image based on the Maximum Entropy principle (ARIEM). From the qualitative analysis of the images, we conclude that the neutron radiographic system was able to detect the explosive sample in a satisfactory way while the quantitative analysis confirmed the application effectiveness of a scanner to acquire the projection dates whose objective is only a reconnaissance. (author)

  15. Radiation-thermal effects change of physico-mechanical properties in reactor materials irradiated with neutrons and energetic charged particles

    International Nuclear Information System (INIS)

    Hofman, A.

    1999-01-01

    In the first part of the report (chapter 1) the earlier results of the important scientific and technological investigations which were performed in the seventies years in Poland have been presented. They concerned the fabrication, corrosion, mechanical properties of materials for research and power reactors. Being of the general survey character, the chapter includes own, original results of research of thermal irradiation effects on microstructure evolution phase transformations and mechanical properties of reactor materials. The kinetics of isothermal transformation β→α in U-Cr 0.4% wt. alloy has been studied. Factors affecting stress-corrosion cracking of zirconium in iodine vapour have been investigated. The rings and loops for irradiation specimens and Hot Laboratory for postirradiation examination of construction materials is described. In the second part (chapters 2, 3, 4, 5) performed the investigations and simulations of radiation damage in metals by heavy ion beams (E > 1 MeV/a.m.n.) were described scientific base and technical problems of the method of irradiation of heavy ions and of the examination of irradiated samples is presented. It is followed by a summary of the results of simulation and reactor experiments on different materials. Radiation hardening of a number metals (Al, Zr, Cu, Ni, U) irradiated by heavy ion and neutrons, mechanical properties and microstructural evolution in ion and neutron irradiated austenitic stainless steel is described. The last chapter is a description of practical aspects of the presented studies in nuclear science and technology. (author)

  16. Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

    Science.gov (United States)

    Basiri, H.; Tavakoli-Anbaran, H.

    2018-01-01

    Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

  17. Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

    Science.gov (United States)

    Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

    2018-01-01

    The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

  18. Thermal conductivity of beryllium under low temperature high dose neutron irradiation

    International Nuclear Information System (INIS)

    Chakin, V.P.; Latypov, R.N.; Suslov, D.N.; Kupriyanov, I.B.

    2004-01-01

    Thermal conductivity of compact beryllium of several Russian grades such as TE-400, TE-56, TE-30, TIP and DIP differing in the production technology, grain size and impurity content has been investigated. The thermal diffusivity of beryllium was measured on the disks in the initial and irradiated conditions using the pulse method in the range from room temperature to 200degC. The thermal conductivity was calculated using the table values for the beryllium thermal capacity. The specimens and beryllium neutron source fragments were irradiation in the SM reactor at 70degC and 200degC to a neutron fluence of (0.5-11.4)·10 22 cm -2 (E>0.1 MeV) and in the BOR-60 reactor at 400degC to 16·10 22 cm -2 (E>0.1MeV), respectively. The low-temperature irradiation leads to the drop decrease of the beryllium thermal conductivity and the effect depends on the irradiation parameters. The paper analyses the effect of irradiation parameters (temperature, neutron fluence), measurement temperature and structural factors on beryllium conductivity. The experiments have revealed that the short time post-irradiation annealing at high temperature results in partial reduction of the thermal conductivity of irradiated beryllium. (author)

  19. Geochemistry of the lunar highlands as revealed by measurements of thermal neutrons.

    Science.gov (United States)

    Peplowski, Patrick N; Beck, Andrew W; Lawrence, David J

    2016-03-01

    Thermal neutron emissions from the lunar surface provide a direct measure of bulk elemental composition that can be used to constrain the chemical properties of near-surface (depth lunar materials. We present a new calibration of the Lunar Prospector thermal neutron map, providing a direct link between measured count rates and bulk elemental composition. The data are used to examine the chemical and mineralogical composition of the lunar surface, with an emphasis on constraining the plagioclase concentration across the highlands. We observe that the regions of lowest neutron absorption, which correspond to estimated plagioclase concentrations of >85%, are generally associated with large impact basins and are colocated with clusters of nearly pure plagioclase identified with spectral reflectance data.

  20. MCT: a Monte Carlo code for time-dependent neutron thermalization problems

    International Nuclear Information System (INIS)

    Cupini, E.; Simonini, R.

    1974-01-01

    In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10 -5 were quickly obtained. (U.S.)

  1. Personal Dosemeter of Thermal Neutron Using A Cr-39 Detector with Filter Natural LiF

    International Nuclear Information System (INIS)

    Sofyan, Hasnel; Thamrin, M.Thoyib

    1996-01-01

    The research of personal dosemeter for thermal neutron using Cr-39 detector with different thicknesses of natural LiF filter was carried out. The irradiation of Cr-39 detector with neutron source from reactor research TRIGA mark II of Rikkyo University Tokyo, Japan. Nuclear track was counted by automatic method with ASPECT ver.4.22 Series A4T124 software and manual method for correction. The result of research, the maximum of nuclear tracks was obtained at 8 mm of LiF filter was 10 mm with 11,630x10E-5 track/neutron for air radiation. And on phantom radiation, the thickness of filter was 10 mm with 11,630x10E-5 track/neutron. Its values were 3.6 and 7.5 bigger than the response of Cr-39 non filter in air and on phantom radiation, respectively

  2. On line local measurement of thermal neutron flux on BNCT patient using SPND

    International Nuclear Information System (INIS)

    Miller, M.E.; Sztejnberg Goncalves-Carralves, M.L.; Gonzalez, S.J.

    2006-01-01

    The first on-line neutron flux measurement on a patient using a self-powered neutron detector (SPND) was assessed during the fourth clinical trial of the Boron Neutron Capture Therapy (BNCT) Project carried out at the National Atomic Energy Commission of Argentina (CNEA) and the medical center Angel H. Roffo. The SPND was specially developed and assembled for BNCT by CNEA. Its small size, 1 cm sensible length and 1.9 mm diameter, allowed performing a localized measurement. Since the treated tumors were cutaneous melanomas of nodular type, the SPND was located on the patient's skin. The patient was exposed to three different and consecutive fields and in each of them the SPND was used to measure local thermal neutron fluxes at selected dosimetric reference points. The values of the measured fluxes agreed with the ones estimated by calculation. This trial also demonstrated the usefulness of the SPND for assessing flux on-line. (author)

  3. Comparative measurements of independent yields of 239Pu fission fragments induced by thermal and resonance neutrons

    International Nuclear Information System (INIS)

    Gundorin, N.A.; Kopach, Y.N.; Telezhnikov, S.A.

    1994-01-01

    The independent yields of 239 Pu fission fragments by means of gamma-spectroscopy method were measured for light and heavy groups on the IBR-30 reactor in Dubna. Comparative analysis of experimental data for fission induced by thermal and resonance neutrons was performed. The possibilities to increase the measurement's precision consist of the employment of a HPGe detector with high efficiency and its open-quotes activeclose quotes shielding in the gamma spectrometer, as well as a high speed electronics system. In this way the number of identified fragments will be increased and independent yields will be measured to a precision of 1-3%. Measurements at the source with shorter neutron pulse duration to increase neutron energy resolution will be possible after the reconstruction of a modern neutron source in Dubna in accordance with the IREN project

  4. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Abou Taleb, W.M. [Alexandria Univ. (Egypt); Madi, N.K.; Kassem, M.E.; El-Khatib, A.M. [Alexandria Univ. (Egypt). Dept. of Physics

    1996-05-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10{sup 9} n/cm{sup 2}. The optical energy gap E{sub op} exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C{sub p} showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author).

  5. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    International Nuclear Information System (INIS)

    Abou Taleb, W.M.; Madi, N.K.; Kassem, M.E.; El-Khatib, A.M.

    1996-01-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10 9 n/cm 2 . The optical energy gap E op exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C p showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author)

  6. Thermal neutron detection using a silicon pad detector and {sup 6}LiF removable converters

    Energy Technology Data Exchange (ETDEWEB)

    Barbagallo, Massimo [Istituto Nazionale di Fisica Nucleare, Sezione di Bari (Italy); Cosentino, Luigi; Marchetta, Carmelo; Pappalardo, Alfio; Scire, Carlotta; Scire, Sergio; Schillaci, Maria; Vecchio, Gianfranco; Finocchiaro, Paolo [Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali del Sud, Catania (Italy); Forcina, Vittorio; Peerani, Paolo [European Commission, Joint Research Centre, Institute of Transuranium Elements, Ispra (Italy); Vaccaro, Stefano [European Commission, Directorate-General for Energy (Luxembourg)

    2013-03-15

    A semiconductor detector coupled with a neutron converter is a good candidate for neutron detection, especially for its compactness and reliability if compared with other devices, such as {sup 3}He tubes, even though its intrinsic efficiency is rather lower. In this paper we show a neutron detector design consisting of a 3 cm Multiplication-Sign 3 cm silicon pad detector coupled with one or two external {sup 6}LiF layers, enriched in {sup 6}Li at 95%, placed in contact with the Si active surfaces. This prototype, first characterized and tested at INFN Laboratori Nazionali del Sud and then at JRC Ispra, was successfully shown to detect thermal neutrons with the expected efficiency and an outstanding gamma rejection capability.

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

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, L G; Moeller, E

    1968-01-15

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

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

    International Nuclear Information System (INIS)

    Larsson, L.G.; Moeller, E.

    1968-01-01

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

  9. Measurement of thermal neutron distribution of flux in the fuel element cluster - Progress report

    International Nuclear Information System (INIS)

    Krcevinac, S.B.; Takac, S.M.

    1966-12-01

    Relative distribution of thermal neutron flux in the fuel element cluster (19 UO 2 rods with Zr-II cladding, D 2 O moderator and coolant) was measured by newly developed cell perturbation method. The obtained values of mean density ratios are compared to the results obtained by TER-I code using Amouyal - Benoist model [sr

  10. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 160A

    International Nuclear Information System (INIS)

    Walker, B. J.; Miller, D. T.

    2017-01-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 160A which were measured by the Radiation Measurements Laboratory (RML).

  11. Further development of thermal neutron capture therapy for metastatic and deeply-invasive human malignant melanoma

    International Nuclear Information System (INIS)

    Mishima, Yutaka

    1995-03-01

    This issue is the collection of the papers presented thermal neutron capture therapy for metastatic and deeply-invasive human malignant melanoma. Separate abstracts were prepared for 2 of the papers in this report. The remaining 32 papers were considered outside the subject scope of INIS. (J.P.N.)

  12. Improvement of mungbean by X-ray and thermal neutron irradiation

    International Nuclear Information System (INIS)

    Kwon, S.H.; Oh, J.H.

    1983-01-01

    With the aim of improving yield, resistance to Cercospora leaf spot and pod shattering, mungbean varieties Kyunggi No. 5 and M-317 were irradiated with X-rays and thermal neutrons. High yielding mutant lines are generally characterized by a higher number of pods per plant. Better Cercospora resistance appears often associated with later maturity. Satisfactory shattering resistance was not yet obtained. (author)

  13. Monte Carlo simulations to advance characterisation of landmines by pulsed fast/thermal neutron analysis

    NARCIS (Netherlands)

    Maucec, M.; Rigollet, C.

    The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra,

  14. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 160A

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, D. T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-06-06

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 160A which were measured by the Radiation Measurements Laboratory (RML).

  15. Simulation of Thermal, Neutronic and Radiation Characteristics in Spent Nuclear Fuel and Radwaste Facilities

    International Nuclear Information System (INIS)

    Poskas, P.; Bartkus, G.

    1999-01-01

    The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)

  16. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    International Nuclear Information System (INIS)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-01-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized

  17. {sup 6}LiF oleic acid capped nanoparticles entrapment in siloxanes for thermal neutron detection

    Energy Technology Data Exchange (ETDEWEB)

    Carturan, S., E-mail: sara.carturan@lnl.infn.it; Maggioni, G., E-mail: Gianluigi.maggioni@lnl.infn.it [Department of Physics and Astronomy, University of Padova, Via Marzolo 8, 35100 Padova (Italy); INFN, Laboratori Nazionali di Legnaro, Viale dell’Università 2, 35020 Legnaro (Italy); Marchi, T.; Gramegna, F.; Cinausero, M. [INFN, Laboratori Nazionali di Legnaro, Viale dell’Università 2, 35020 Legnaro (Italy); Quaranta, A. [Department of Industrial Engineering, University of Trento, Trento (Italy); INFN, Tifpa, Trento (Italy); Palma, M. Dalla [INFN, Laboratori Nazionali di Legnaro, Viale dell’Università 2, 35020 Legnaro (Italy); Department of Industrial Engineering, University of Trento, Trento (Italy)

    2016-07-07

    The good light output of siloxane based scintillators as displayed under γ-rays and α particles has been exploited here to obtain clear and reliable response toward thermal neutrons. Sensitization towards thermal neutrons has been pursued by adding {sup 6}LiF, in form of nanoparticles. Aiming at the enhancement of compatibility between the inorganic nanoparticles and the low polarity, siloxane based surrounding medium, oleic acid-capped {sup 6}LiF nanoparticles have been synthesized by thermal decomposition of Li trifluoroacetate. Thin pellets siloxane scintillator maintained their optical transmittance up to weight load of 2% of {sup 6}Li. Thin samples with increasing {sup 6}Li concentration and thicker ones with fixed {sup 6}Li amount have been prepared and tested with several sources (α, γ-rays, moderated neutrons). Light output as high as 80% of EJ212 under α irradiation was measured with thin samples, and negligible changes have been observed as a result of {sup 6}LiF addition. In case of thick samples, severe light loss has been observed, as induced by opacity. Nevertheless, thermal neutrons detection has been assessed and the data have been compared with GS20, based on Li glass, taken as a reference material.

  18. Prompt γ-ray data evaluation of thermal-neutron capture for A = 1-25

    International Nuclear Information System (INIS)

    Zhou Chunmei

    1999-01-01

    The method of prompt γ-ray data evaluation for thermal-neutron capture has been briefly presented. The prompt capture γ-ray data of stable nuclei for A = 1 - 25 are evaluated. The evaluated data have been changed into the ENSDF format and the checks of physics and format have been made

  19. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  20. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  1. Intraoperative boron neutron capture therapy for malignant gliomas. First clinical results of Tsukuba phase I/II trial using JAERI mixed thermal-epithermal beam

    International Nuclear Information System (INIS)

    Matsumura, A.; Yamamoto, T.; Shibata, Y.

    2000-01-01

    Since October 1999, a clinical trial of intraoperative boron neutron capture therapy (IOBNCT) is in progress at JRR-4 (Japan Research Reactor-4) in Japan Atomic Energy Research Institute (JAERI) using mixed thermal-epithermal beam (thermal neutron beam I: TNB-I). Compared to pure thermal beam (thermal neutron beam II: TNB-II), TNB-I has an improved neutron delivery into the deep region than TNB-II. The clinical protocol and the preliminary results will be discussed. (author)

  2. 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.

  3. Superconductivity degradation in Gd-containing high temperature superconductors (HTSC) under thermal neutron irradiation

    International Nuclear Information System (INIS)

    Petrov, A.; Kudrenitskis, I.; Makletsov, A.; Arhipov, A.; Karklin, N.

    1999-01-01

    The physical properties of ordered crystals are extremely sensitive to the degree of order in the distribution of the various kinds of atoms over the corresponding sites in the crystal lattice. An increasingly popular means of creating disordered states is to use nuclear radiation. The type of radiation defects which appear and the nature and degree of the structural changes in ordered crystals depend on the kind of radiation and the fluence level, the irradiation temperature, the type of crystal structure, the composition and initial disorder of the material, the character of the interatomic forces, etc. There are many such scientific publications where the effects of fast neutron irradiation on high temperature superconductors (HTSC) have been studied in both polycrystalline and single crystalline superconductors. It is known also that the role of thermal neutrons in structural defects forming is negligible in comparison with fast neutrons because of their small (∼0.025 eV) energy. But it is evident enough that in superconductors containing isotopes with large thermal neutron cross sections the important results concerning the role of point defects could be obtained. Such point defects are creating due to soft displacements of isotopes having interacted with thermal neutrons. Such the possibility of creating point defects in solids including HTSC is investigating by several groups (Austria, USA, China, Latvia) and these investigations have found the support in the person of IAEA. In this review the authors consider the changes brought about by thermal-neutron irradiation (E∼0.025 eV) in the structure, superconducting and magnetic properties of gadolinium containing ordered HTSC with the structure 123, whose extreme electric and magnetic properties continue to attract both research and practical interest. All of the studies reviewed have been done on bulk polycrystalline samples RBa 2 Cu 3 O 7-δ (where R - natural mixture of Gd isotopes, 155 Gd, 157 Gd, 160

  4. Simultaneous measurement of neutrons and fission fragments of thermal neutron fission of U-233

    International Nuclear Information System (INIS)

    Itsuro Kimura; Katsuhisa Nishio; Yoshihiro Nakagome

    2000-01-01

    The multiplicity and the energy of prompt neutrons from the fragments for 233 U(n th , f) were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 u. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 u. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1 MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppresses, the total excitation energy is significantly large and reaches to about 40 MeV, suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point [ru

  5. Thermal Evolution and Crystallisation Regimes of the Martian Core

    Science.gov (United States)

    Davies, C. J.; Pommier, A.

    2015-12-01

    Though it is accepted that Mars has a sulfur-rich metallic core, its chemical and physical state as well as its time-evolution are still unconstrained and debated. Several lines of evidence indicate that an internal magnetic field was once generated on Mars and that this field decayed around 3.7-4.0 Gyrs ago. The standard model assumes that this field was produced by a thermal (and perhaps chemical) dynamo operating in the Martian core. We use this information to construct parameterized models of the Martian dynamo in order to place constraints on the thermochemical evolution of the Martian core, with particular focus on its crystallization regime. Considered compositions are in the FeS system, with S content ranging from ~10 and 16 wt%. Core radius, density and CMB pressure are varied within the errors provided by recent internal structure models that satisfy the available geodetic constraints (planetary mass, moment of inertia and tidal Love number). We also vary the melting curve and adiabat, CMB heat flow and thermal conductivity. Successful models are those that match the dynamo cessation time and fall within the bounds on present-day CMB temperature. The resulting suite of over 500 models suggest three possible crystallization regimes: growth of a solid inner core starting at the center of the planet; freezing and precipitation of solid iron (Fe- snow) from the core-mantle boundary (CMB); and freezing that begins midway through the core. Our analysis focuses on the effects of core properties that are expected to be constrained during the forthcoming Insight mission.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-11

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  9. Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

    International Nuclear Information System (INIS)

    Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

    2014-01-01

    Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

  10. Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

    Science.gov (United States)

    Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

    2006-02-01

    GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

  11. Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki, E-mail: koyanagit@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kumar, N.A.P. Kiran [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hwang, Taehyun [Tohoku University, Sendai, 980-8579 (Japan); Garrison, Lauren M.; Hu, Xunxiang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2017-07-15

    Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

  12. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  13. Relative measurement of the fluxes of thermal, resonant and rapid neutrons in reactor G1

    International Nuclear Information System (INIS)

    Carle, R.; Mazancourt, T. de

    1957-01-01

    We sought to determine the behavior of the thermal, resonant and rapid neutron fluxes in the multiplier-reflector transition region, in the two principal directions of the system. We have also measured the variation of these different fluxes in the body of the multiplier medium in a canal filled with graphite and in an empty canal. The results are given in the form of curves representing: - the variation of the ratio of the thermal flux to the rapid flux in axial and radial transitions - the behavior of the thermal and resonant fluxes and the variation of their ratio in the same regions. (author) [fr

  14. Low temperature thermal annealing in fast neutron-irradiated potassium permanganate

    Energy Technology Data Exchange (ETDEWEB)

    Owens, C W; Lecington, W C [New Hampshire Univ., Durham (USA). Dept. of Chemistry

    1975-01-01

    The effect of thermal annealing on the retention of recoil /sup 54/Mn as permanganate in crystalline KMnO/sub 4/ irradiated with fast neutrons at liquid nitrogen temperature has been studied. The retention after 4 hrs of annealing increases from about 8% at -196/sup 0/ to a maximum of 61% at 180/sup 0/, then decreases at higher temperatures. A single activation energy (approximately 0.01 eV) applies to the thermal annealing process between -196/sup 0/ and -40/sup 0/. Extrapolation of the data suggests that below -229/sup 0/ no thermal annealing would occur.

  15. The role of neutron star mergers in the chemical evolution of the Galactic halo

    Science.gov (United States)

    Cescutti, G.; Romano, D.; Matteucci, F.; Chiappini, C.; Hirschi, R.

    2015-05-01

    Context. The dominant astrophysical production site of the r-process elements has not yet been unambiguously identified. The suggested main r-process sites are core-collapse supernovae and merging neutron stars. Aims: We explore the problem of the production site of Eu. We also use the information present in the observed spread in the Eu abundances in the early Galaxy, and not only its average trend. Moreover, we extend our investigations to other heavy elements (Ba, Sr, Rb, Zr) to provide additional constraints on our results. Methods: We adopt a stochastic chemical evolution model that takes inhomogeneous mixing into account. The adopted yields of Eu from merging neutron stars and from core-collapse supernovae are those that are able to explain the average [Eu/Fe]-[Fe/H] trend observed for solar neighbourhood stars, the solar abundance of Eu, and the present-day abundance gradient of Eu along the Galactic disc in the framework of a well-tested homogeneous model for the chemical evolution of the Milky Way. Rb, Sr, Zr, and Ba are produced by both the s- and r-processes. The r-process yields were obtained by scaling the Eu yields described above according to the abundance ratios observed in r-process rich stars. The s-process contribution by spinstars is the same as in our previous papers. Results: Neutron star binaries that merge in less than 10 Myr or neutron star mergers combined with a source of r-process generated by massive stars can explain the spread of [Eu/Fe] in the Galactic halo. The combination of r-process production by neutron star mergers and s-process production by spinstars is able to reproduce the available observational data for Sr, Zr, and Ba. We also show the first predictions for Rb in the Galactic halo. Conclusions: We confirm previous results that either neutron star mergers on a very short timescale or both neutron star mergers and at least a fraction of Type II supernovae have contributed to the synthesis of Eu in the Galaxy. The r

  16. Visualization of thermal management system in space using neutron radiography

    International Nuclear Information System (INIS)

    Nakazawa, Takeshi

    1995-01-01

    The visualizing technique by neutron radiography is effective for visualizing liquid in metals, and the applications in wide fields have been reported. In this paper, as one of the examples of applying the visualizing technique by neutron radiography, the experiment of visualizing the two-phase fluid loop heat removal system for the purpose of using in spatial environment was carried out, and its results are reported. For future large scale space ships and space stations, the heat removal system with two-phase fluid loop which utilizes the phase transformation of heat transport media is regarded as promising. By this system, good heat transfer performance is obtained, transported heat quantity per unit mass of media increases, and pumping power and the weight of the total system are reduced. Temperature can be controlled by system pressure. The two-phase fluid loop for the visualization experiment and the experimental results are reported. By the experiment using the real time NRG system at the JRR-3M, the boiling and evaporation phenomena in the capillary heat transfer tubes were able to be visualized. (K.I.)

  17. Systematic thermal reduction of neutronization in core-collapse supernovae

    International Nuclear Information System (INIS)

    Fantina, A.F.; Donati, P.; Pizzochero, P.M.

    2009-01-01

    We investigate to what extent the temperature dependence of the nuclear symmetry energy can affect the neutronization of the stellar core prior to neutrino trapping during gravitational collapse. To this end, we implement a one-zone simulation to follow the collapse until β-equilibrium is reached and the lepton fraction remains constant. Since the strength of electron capture on the neutron-rich nuclei associated to the supernova scenario is still an open issue, we keep it as a free parameter. We find that the temperature dependence of the symmetry energy consistently yields a small reduction of deleptonization, which corresponds to a systematic effect on the shock wave energetics: the gain in dissociation energy of the shock has a small yet non-negligible value of about 0.4 foe (1 foe=10 51 erg) and this result is almost independent from the strength of nuclear electron capture. The presence of such a systematic effect and its robustness under changes of the parameters of the one-zone model are significant enough to justify further investigations with detailed numerical simulations of supernova explosions.

  18. Compressive strength evolution of thermally-stressed Saint Maximin limestone.

    Science.gov (United States)

    Farquharson, J.; Griffiths, L.; Baud, P.; Wadsworth, F. B.; Heap, M. J.

    2017-12-01

    The Saint Maximin quarry (Oise, France) opened in the early 1600s, and its limestone has been used extensively as masonry stone, particularly during the classical era of Parisian architecture from the 17th century onwards. Its widespread use has been due to a combination of its regional availability, its high workability, and its aesthetic appeal. Notable buildings completed using this material include sections of the Place de la Concorde and the Louvre in Paris. More recently, however, it has seen increasing use in the construction of large private residences throughout the United States as well as extensions to private institutions such as Stanford University. For any large building, fire hazard can be a substantial concern, especially in tectonically active areas where catastrophic fires may arise following large-magnitude earthquakes. Typically, house fires burn at temperatures of around 600 °C ( 1000 F). Given the ubiquity of this geomaterial as a building stone, it is important to ascertain the influence of heating on the strength of Saint Maximin limestone (SML), and in turn the structural stability of the buildings it is used in. We performed a series of compressive tests and permeability measurements on samples of SML to determine its strength evolution in response to heating to incrementally higher temperatures. We observe that the uniaxial compressive strength of SML decreases from >12 MPa at room temperature to 400 °C). We anticipate that this substantial weakening is in part a result of thermal microcracking, whereby changes in temperature induce thermal stresses due to a mismatch in thermal expansion between the constituent grains. This mechanism is compounded by the volumetric increase of quartz through its alpha - beta transition at 573 °C, and by the thermal decomposition of calcite. To track the formation of thermal microcracks, we monitor acoustic emissions, a common proxy for microcracking, during the heating of an SML sample. The

  19. Removal, transportation and disposal of the Millstone 2 neutron thermal shield

    International Nuclear Information System (INIS)

    Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.

    1985-01-01

    Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility

  20. Prompt gamma-ray analysis using JRR-3M cold and thermal neutron guide beams

    International Nuclear Information System (INIS)

    Yonezawa, C.; Haji Wood, A.K.; Magara, M.; Hoshi, M.; Tachikawa, E.; Sawahata, H.; Ito, Y.

    1993-01-01

    A permanent and stand-alone neutron-induced prompt gamma-ray analysis (PGA) system, usable at both cold and thermal neutron beam guides of JRR-3M has been constructed. Neutron flux at the sample positions were 1.4x10 8 and 2.4x10 7 n cm -2 s -1 for the cold and thermal neutrons, respectively. The γ-ray spectrometer is equipped to acquire three modes of spectra simultaneously: single mode, Compton suppression mode and pair mode, in an energy range up to 12 MeV. Owing to the cold neutron guide beam and the low γ-ray background system, analytical sensitivities and detection limits better than those in other PGA systems have been achieved. Analytical sensitivity and detection limit for 73 elements were measured. Boron, Gd, Sm and Cd are the most sensitive elements with detection limits down to 1 to 10 ng. For some elements such as F, Al, V, Eu and Hf, decay γ-rays are more sensitive compared to their respective prompt γ-ray. Analytical sensitivity of several heavy elements through detection of characteristic X-rays was higher than that through the prompt γ-ray detection. Analytical applicability of some sensitive elements such as B, H, Gd and Sm were examined. Isotopic analysis of Ni and Si were also examined. (author)