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Sample records for neutron dose yields

  1. Neutron yields and effective doses produced by Galactic Cosmic Ray interactions in shielded environments in space

    NASA Astrophysics Data System (ADS)

    Heilbronn, Lawrence H.; Borak, Thomas B.; Townsend, Lawrence W.; Tsai, Pi-En; Burnham, Chelsea A.; McBeth, Rafe A.

    2015-11-01

    In order to define the ranges of relevant neutron energies for the purposes of measurement and dosimetry in space, we have performed a series of Monte Carlo transport model calculations that predict the neutron field created by Galactic Cosmic Ray interactions inside a variety of simple shielding configurations. These predictions indicate that a significant fraction of the neutron fluence and neutron effective dose lies in the region above 20 MeV up to several hundred MeV. These results are consistent over thicknesses of shielding that range from very thin (2.7 g/cm2) to thick (54 g/cm2), and over both shielding materials considered (aluminum and water). In addition to these results, we have also investigated whether simplified Galactic Cosmic Ray source terms can yield predictions that are equivalent to simulations run with a full GCR source term. We found that a source using a GCR proton and helium spectrum together with a scaled oxygen spectrum yielded nearly identical results to a full GCR spectrum, and that the scaling factor used for the oxygen spectrum was independent of shielding material and thickness. Good results were also obtained using a GCR proton spectrum together with a scaled helium spectrum, with the helium scaling factor also independent of shielding material and thickness. Using a proton spectrum alone was unable to reproduce the full GCR results.

  2. Dose estimations of fast neutrons from a nuclear reactor by micronuclear yields in onion seedlings.

    PubMed

    Fujikawa, K; Endo, S; Itoh, T; Yonezawa, Y; Hoshi, M

    1999-12-01

    Irradiations of onion seedlings with fission neutrons from bare, Pb-moderated, and Fe-moderated 252Cf sources induced micronuclei in the root-tip cells at similar rates. The rate per cGy averaged for the three sources, , was 19 times higher than rate induced by 60Co gamma-rays. When neutron doses, Dn, were estimated from frequencies of micronuclei induced in onion seedlings after exposure to neutron-gamma mixed radiation from a 1 W nuclear reactor, using the reciprocal of as conversion factor, resulting Dn values agreed within 10% with doses measured with paired ionizing chambers. This excellent agreement was achieved by the high sensitivity of the onion system to fast neutrons relative to gamma-rays and the high contribution of fast neutrons to the total dose of mixed radiation in the reactor's field.

  3. Dependence of TLD thermoluminescence yield on absorbed dose in a thermal neutron field.

    PubMed

    Gambarini, G; Roy, M S

    1997-01-01

    The emission from 6LiF and 7LiF thermoluminescence dosimeters (TLDs) exposed to the mixed field of thermal neutrons and gamma-rays of the thermal facility of a TRIGA MARK II nuclear reactor has been investigated for various thermal neutron fluences of the order of magnitude of those utilised in radiotherapy, with the purpose of investigating the reliability of TLD readouts in such radiation fields and of giving some information for better obtainment of the absorbed dose values. The emission after exposure in this mixed field is compared with the emission after gamma-rays only. The glow curves have been deconvoluted into gaussian peaks, and the differences in the characteristics of the peaks observed for the two radiation fields, having different linear energy transfers, and for different doses are shown. Irreversible radiation damage in dosimeters having high sensitivity to thermal neutrons is also reported, showing a memory effect of the previous thermal neutron irradiation history which is not restored by anneal treatment. PMID:9463872

  4. Radiolytic yield of ozone in air for low dose neutron and x-ray/gamma-ray radiation

    NASA Astrophysics Data System (ADS)

    Cole, J.; Su, S.; Blakeley, R. E.; Koonath, P.; Hecht, A. A.

    2015-01-01

    Radiation ionizes surrounding air and produces molecular species, and these localized effects may be used as a signature of, and for quantification of, radiation. Low-level ozone production measurements from radioactive sources have been performed in this work to understand radiation chemical yields at low doses. The University of New Mexico AGN-201 M reactor was used as a tunable radiation source. Ozone levels were compared between reactor-on and reactor-off conditions, and differences (0.61 to 0.73 ppb) well below background levels were measured. Simulations were performed to determine the dose rate distribution and average dose rate to the air sample within the reactor, giving 35 mGy of mixed photon and neutron dose. A radiation chemical yield for ozone of 6.5±0.8 molecules/100 eV was found by a variance weighted average of the data. The different contributions of photons and neutrons to radiolytic ozone production are discussed.

  5. Neutron dose equivalent meter

    DOEpatents

    Olsher, Richard H.; Hsu, Hsiao-Hua; Casson, William H.; Vasilik, Dennis G.; Kleck, Jeffrey H.; Beverding, Anthony

    1996-01-01

    A neutron dose equivalent detector for measuring neutron dose capable of accurately responding to neutron energies according to published fluence to dose curves. The neutron dose equivalent meter has an inner sphere of polyethylene, with a middle shell overlying the inner sphere, the middle shell comprising RTV.RTM. silicone (organosiloxane) loaded with boron. An outer shell overlies the middle shell and comprises polyethylene loaded with tungsten. The neutron dose equivalent meter defines a channel through the outer shell, the middle shell, and the inner sphere for accepting a neutron counter tube. The outer shell is loaded with tungsten to provide neutron generation, increasing the neutron dose equivalent meter's response sensitivity above 8 MeV.

  6. Dose equivalent neutron dosimeter

    DOEpatents

    Griffith, Richard V.; Hankins, Dale E.; Tomasino, Luigi; Gomaa, Mohamed A. M.

    1983-01-01

    A neutron dosimeter is disclosed which provides a single measurements indicating the amount of potential biological damage resulting from the neutron exposure of the wearer, for a wide range of neutron energies. The dosimeter includes a detecting sheet of track etch detecting material such as a carbonate plastic, for detecting higher energy neutrons, and a radiator layer containing conversion material such as .sup.6 Li and .sup.10 B lying adjacent to the detecting sheet for converting moderate energy neutrons to alpha particles that produce tracks in the adjacent detecting sheet. The density of conversion material in the radiator layer is of an amount which is chosen so that the density of tracks produced in the detecting sheet is proportional to the biological damage done by neutrons, regardless of whether the tracks are produced as the result of moderate energy neutrons striking the radiator layer or as the result of higher energy neutrons striking the sheet of track etch material.

  7. Photon doses in NPL standard neutron fields.

    PubMed

    Roberts, N J; Horwood, N A; McKay, C J

    2014-10-01

    Standard neutron fields are invariably accompanied by a photon component due to the neutron-generating reactions and secondary neutron interactions in the surrounding environment. A set of energy-compensated Geiger-Müller (GM) tubes and electronic personal dosemeters (EPDs) have been used to measure the photon dose rates in a number of standard radionuclide and accelerator-based neutron fields. The GM tubes were first characterised in standard radioisotope and X-ray photon fields and then modelled using MCNP to determine their photon dose response as a function of energy. Values for the photon-to-neutron dose equivalent ratios are presented and compared with other published values.

  8. Variation in lunar neutron dose estimates.

    PubMed

    Slaba, Tony C; Blattnig, Steve R; Clowdsley, Martha S

    2011-12-01

    The radiation environment on the Moon includes albedo neutrons produced by primary particles interacting with the lunar surface. In this work, HZETRN2010 is used to calculate the albedo neutron contribution to effective dose as a function of shielding thickness for four different space radiation environments and to determine to what extent various factors affect such estimates. First, albedo neutron spectra computed with HZETRN2010 are compared to Monte Carlo results in various radiation environments. Next, the impact of lunar regolith composition on the albedo neutron spectrum is examined, and the variation on effective dose caused by neutron fluence-to-effective dose conversion coefficients is studied. A methodology for computing effective dose in detailed human phantoms using HZETRN2010 is also discussed and compared. Finally, the combined variation caused by environmental models, shielding materials, shielding thickness, regolith composition and conversion coefficients on the albedo neutron contribution to effective dose is determined. It is shown that a single percentage number for characterizing the albedo neutron contribution to effective dose can be misleading. In general, the albedo neutron contribution to effective dose is found to vary between 1-32%, with the environmental model, shielding material and shielding thickness being the driving factors that determine the exact contribution. It is also shown that polyethylene or other hydrogen-rich materials may be used to mitigate the albedo neutron exposure. PMID:21859325

  9. Radiation Dose from Lunar Neutron Albedo

    NASA Technical Reports Server (NTRS)

    Adams, J. H., Jr.; Bhattacharya, M.; Lin, Zi-Wei; Pendleton, G.

    2006-01-01

    The lunar neutron albedo from thermal energies to 8 MeV was measured on the Lunar Prospector Mission in 1998-1999. Using GEANT4 we have calculated the neutron albedo due to cosmic ray bombardment of the moon and found a good-agreement with the measured fast neutron spectra. We then calculated the total effective dose from neutron albedo of all energies, and made comparisons with the effective dose contributions from both galactic cosmic rays and solar particle events to be expected on the lunar surface.

  10. Dose-equivalent neutron dosimeter

    DOEpatents

    Griffith, R.V.; Hankins, D.E.; Tomasino, L.; Gomaa, M.A.M.

    1981-01-07

    A neutron dosimeter is disclosed which provides a single measurement indicating the amount of potential biological damage resulting from the neutron exposure of the wearer, for a wide range of neutron energies. The dosimeter includes a detecting sheet of track etch detecting material such as a carbonate plastic, for detecting higher energy neutrons, and a radiator layer contaning conversion material such as /sup 6/Li and /sup 10/B lying adjacent to the detecting sheet for converting moderate energy neutrons to alpha particles that produce tracks in the adjacent detecting sheet.

  11. Dose spectra from energetic particles and neutrons

    NASA Astrophysics Data System (ADS)

    Schwadron, Nathan; Bancroft, Chris; Bloser, Peter; Legere, Jason; Ryan, James; Smith, Sonya; Spence, Harlan; Mazur, Joe; Zeitlin, Cary

    2013-10-01

    spectra from energetic particles and neutrons (DoSEN) are an early-stage space technology research project that combines two advanced complementary radiation detection concepts with fundamental advantages over traditional dosimetry. DoSEN measures not only the energy but also the charge distribution (including neutrons) of energetic particles that affect human (and robotic) health in a way not presently possible with current dosimeters. For heavy ions and protons, DoSEN provides a direct measurement of the lineal energy transfer (LET) spectra behind shielding material. For LET measurements, DoSEN contains stacks of thin-thick Si detectors similar in design to those used for the Cosmic Ray Telescope for the Effects of Radiation. With LET spectra, we can now directly break down the observed spectrum of radiation into its constituent heavy-ion components and through biologically based quality factors that provide not only doses and dose rates but also dose equivalents, associated rates, and even organ doses. DoSEN also measures neutrons from 10 to 100 MeV, which requires enough sensitive mass to fully absorb recoil particles that the neutrons produce. DoSEN develops the new concept of combining these independent measurements and using the coincidence of LET measurements and neutron detection to significantly reduce backgrounds in each measurement. The background suppression through the use of coincidence allows for significant reductions in size, mass, and power needed to provide measurements of dose, neutron dose, dose equivalents, LET spectra, and organ doses. Thus, we introduce the DoSEN concept: a promising low-mass instrument that detects the full spectrum of energetic particles, heavy ions, and neutrons to determine biological impact of radiation in space.

  12. Low dose neutron late effects: Cataractogenesis

    SciTech Connect

    Worgul, B.V.

    1991-12-01

    The work is formulated to resolve the uncertainty regarding the relative biological effectiveness (RBE) of low dose neutron radiation. The study exploits the fact that cataractogenesis is sensitive to the inverse dose-rate effect as has been observed with heavy ions and was an endpoint considered in the follow-up of the A-bomb survivors. The neutron radiations were initiated at the Radiological Research Accelerator facility (RARAF) of the Nevis Laboratory of Columbia University. Four week old ({plus minus} 1 day) rats were divided into eight dose groups each receiving single or fractionated total doses of 0.2, 1.0, 5.0 and 25.0 cGy of monoenergetic 435 KeV neutrons. Special restraining jigs insured that the eye, at the midpoint of the lens, received the appropriate energy and dose with a relative error of {plus minus}5%. The fractionation regimen consisted of four exposures, each administered at three hour ({plus minus}) intervals. The neutron irradiated groups are being compared to rats irradiated with 250kVp X-rays in doses ranging from 0.5 to 7 Gy. The animals are being examined on a biweekly basis utilizing conventional slit-lamp biomicroscopy and the Scheimpflug Slit Lamp Imaging System (Zeiss). The follows-ups, entering their second year, will continue throughout the life-span of the animals. This is essential inasmuch as given the extremely low doses which are being utilized clinically detectable opacities were not anticipated until a significant fraction of the life span has lapsed. Current data support this contention. At this juncture cataracts in the irradiated groups are beginning to exceed control levels.

  13. Multigroup neutron dose calculations for proton therapy

    SciTech Connect

    Kelsey Iv, Charles T; Prinja, Anil K

    2009-01-01

    We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations.

  14. Dose Calibration of the ISS-RAD Fast Neutron Detector

    NASA Technical Reports Server (NTRS)

    Zeitlin, C.

    2015-01-01

    The ISS-RAD instrument has been fabricated by Southwest Research Institute and delivered to NASA for flight to the ISS in late 2015 or early 2016. ISS-RAD is essentially two instruments that share a common interface to ISS. The two instruments are the Charged Particle Detector (CPD), which is very similar to the MSL-RAD detector on Mars, and the Fast Neutron Detector (FND), which is a boron-loaded plastic scintillator with readout optimized for the 0.5 to 10 MeV energy range. As the FND is completely new, it has been necessary to develop methodology to allow it to be used to measure the neutron dose and dose equivalent. This talk will focus on the methods developed and their implementation using calibration data obtained in quasi-monoenergetic (QMN) neutron fields at the PTB facility in Braunschweig, Germany. The QMN data allow us to determine an approximate response function, from which we estimate dose and dose equivalent contributions per detected neutron as a function of the pulse height. We refer to these as the "pSv per count" curves for dose equivalent and the "pGy per count" curves for dose. The FND is required to provide a dose equivalent measurement with an accuracy of ?10% of the known value in a calibrated AmBe field. Four variants of the analysis method were developed, corresponding to two different approximations of the pSv per count curve, and two different implementations, one for real-time analysis onboard ISS and one for ground analysis. We will show that the preferred method, when applied in either real-time or ground analysis, yields good accuracy for the AmBe field. We find that the real-time algorithm is more susceptible to chance-coincidence background than is the algorithm used in ground analysis, so that the best estimates will come from the latter.

  15. Morphological transformation of Syrian hamster embryo cells by low doses of fission neutrons delivered at different dose rates

    SciTech Connect

    Jones, C.A.; Sedita, B.A. ); Hill, C.K. . Cancer Research Lab.); Elkind, M.M. . Dept. of Radiology and Radiation Biology)

    1991-01-01

    Both induction of cell transformation and killing were examined with Syrian hamster embryo (SHE) fibroblasts exposed to low doses of JANUS fission-spectrum neutrons delivered at high (10.3 cGy/min) and low (0.43 and 0.086 cGy/min) dose rates. Second-passage cells were irradiated in mass cultures, then cloned over feeder cells. Morphologically transformed colonies were identified 8-10 days later. Cell killing was independent of dose rate, but the yield of transformation was greater after low-dose-rate irradiations. Decreasing the neutron dose-rate from 10.3 to 0.086 cGy/min resulted in a two- to threefold increase in the yield of transformation for neutron exposures below 50 cGy, and enhancement which was consistently observed in repetitive experiments in different radiosensitive SHE cell preparations. 43 refs., 5 figs., 1 tab.

  16. Neutron Yield Measurements via Aluminum Activation

    SciTech Connect

    1999-12-08

    Neutron activation of aluminum may occur by several neutron capture reactions. Four such reactions are described here: {sup 27}Al + n = {sup 28}Al, {sup 27}Al(n,{alpha}){sup 24}Na, {sup 27}Al(n, 2n){sup 26}Al and {sup 27}Al(n,p){sup 27}Mg. The radioactive nuclei {sup 28}Al, {sup 24}Na, and {sup 27}Mg, which are produced via the {sup 27}Al + n = {sup 28}Al, {sup 27}Al(n,{alpha}){sup 24}Na and {sup 27}Al(n,p){sup 27}Mg neutron reactions, beta decay to excited states of {sup 28}Si, {sup 24}Mg and {sup 27}Al respectively. These excited states then emit gamma rays as the nuclei de-excite to their respective ground states.

  17. Low dose neutron late effects: Cataractogenesis

    SciTech Connect

    Worgul, B.V.

    1991-04-01

    The work is formulated to resolve the uncertainty regarding the relative biological effectiveness. The endpoint which is being utilized is cataractogenesis. The advantages conferred by this system stems primarily from the non-invasive longitudinal analysis which it allows. It also exploits a well defined system and one which has demonstrated sensitivity to the inverse dose rate effect observed with heavy ions. Four week old rats were divided into 8 dose groups which received single or fractionated total doses of .2, 1.0, 5.0 and 25 cGy of monoenergetic 435 keV neutrons. Special restraining jigs were devised to insure that the eye at the midpoint of the lens received the appropriate energy and dose with a relative error of {plus minus} 5%. The fractionated regimen consisted of four exposures, each administered at 3 hour intervals. The reference radiations, 250 kVp X-rays, were administered in the same fashion but in doses ranging from .5 to 6.0 Gy. The animals are examined on a bi-weekly basis utilizing conventional slit-lamp biomicroscopy and the Scheimpflug Slit-lamp Imaging System. The follow-ups will continue throughout the lifespan of the animals. When opacification begins full documentation will involve the Zeiss imaging system and Oxford retroillumination photography. The processing routinely employs the Merriam/Focht scoring system for cross-referencing with previous cataract studies and establish cataractogenecity using a proven scoring method.

  18. Measurement of delayed-neutron yield from {sup 237}Np fission induced by thermal neutrons

    SciTech Connect

    Gundorin, N. A.; Zhdanova, K. V.; Zhuchko, V. E.; Pikelner, L. B. Rebrova, N. V.; Salamatin, I. M.; Smirnov, V. I.; Furman, V. I.

    2007-06-15

    The delayed-neutron yield from thermal-neutron-induced fission of the {sup 237}Np nucleus was measured using a sample periodically exposed to a pulsed neutron beam with subsequent detection of neutrons during the time intervals between pulses. The experiment was realized on an Isomer-M setup mounted in the IBR-2 pulsed reactor channel equipped with a mirror neutron guide. The setup and the experimental procedure are described, the background sources are thoroughly analyzed, and the experimental data are presented. The total delayed-neutron yield from {sup 237}Np fission induced by thermal neutrons is {nu}{sub d} = 0.0110 {+-} 0.0009. This study was performed at the Frank Laboratory of Neutron Physics (JINR, Dubna)

  19. Determination of neutron absorbed doses in lithium aluminates.

    PubMed

    Delfín Loya, A; Carrera, L M; Ureña-Núñez, F; Palacios, O; Bosch, P

    2003-04-01

    Lithium-based ceramics have been proposed as tritium breeders for fusion reactors. The lithium aluminate (gamma phase) seems to be thermally and structurally stable, the damages produced by neutron irradiation depend on the absorbed dose. A method based on the measurement of neutron activation of foils through neutron capture has been developed to obtain the neutron absorbed dose in lithium aluminates irradiated in the thermal column facility and in the fixed irradiation system of a Triga Mark III Nuclear Reactor. PMID:12672632

  20. Tensile property changes of metals irradiated to low doses with fission, fusion and spallation neutrons

    SciTech Connect

    Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

    1991-11-01

    Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36--55{degrees}C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90{degrees}C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa.

  1. Neutron fluence-to-dose conversion coefficients for embryo and fetus.

    PubMed

    Chen, Jing; Meyerhof, Dorothy; Vlahovich, Slavica

    2004-01-01

    A problem of concern in radiation protection is the exposure of pregnant women to ionising radiation, because of the high radiosensitivity of the embryo and fetus. External neutron exposure is of concern when pregnant women travel by aeroplane. Dose assessments for neutrons frequently rely on fluence-to-dose conversion coefficients. While neutron fluence-to-dose conversion coefficients for adults are recommended in International Commission on Radiological Protection publications and International Commission on Radiological Units and Measurements reports, conversion coefficients for embryos and fetuses are not given in the publications. This study undertakes Monte Carlo calculations to determine the mean absorbed doses to the embryo and fetus when the mother is exposed to neutron fields. A new set of mathematical models for the embryo and fetus has been developed at Health Canada and is used together with mathematical phantoms of a pregnant female developed at Oak Ridge National Laboratory. Monoenergetic neutrons from 1 eV to 10 MeV are considered in this study. The irradiation geometries include antero-posterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT) and isotropic (ISO) geometries. At each of these standard irradiation geometries, absorbed doses to the fetal brain and body are calculated; for the embryo at 8 weeks and the fetus at 3, 6 or 9 months. Neutron fluence-to-absorbed dose conversion coefficients are derived for the four age groups. Neutron fluence-to-equivalent dose conversion coefficients are given for the AP irradiations which yield the highest radiation dose to the fetal body in the neutron energy range considered here. The results indicate that for neutrons <10 MeV more protection should be given to pregnant women in the first trimester due to the higher absorbed dose per unit neutron fluence to the fetus. PMID:15353732

  2. SU-E-T-602: Beryllium Seeds Implant for Photo-Neutron Yield Using External Beam Therapy

    SciTech Connect

    Koren, S; Veltchev, I; Furhang, E

    2014-06-01

    Purpose: To evaluate the Neutron yield obtained during prostate external beam irradiation. Methods: Neutrons, that are commonly a radiation safety concern for photon beams with energy above 10 MV, are induced inside a PTV from Beryllium implemented seeds. A high megavoltage photon beam delivered to a prostate will yield neutrons via the reaction Be-9(γ,n)2?. Beryllium was chosen for its low gamma,n reaction cross-section threshold (1.67 MeV) to be combined with a high feasible 25 MV photon beam. This beam spectra has a most probable photon energy of 2.5 to 3.0 MeV and an average photon energy of about 5.8 MeV. For this feasibility study we simulated a Beryllium-made common seed dimension (0.1 cm diameter and 0.5 cm height) without taking into account encapsulation. We created a 0.5 cm grid loading pattern excluding the Urethra, using Variseed (Varian inc.) A total of 156 seeds were exported to a 4cm diameter prostate sphere, created in Fluka, a particle transport Monte Carlo Code. Two opposed 25 MV beams were simulated. The evaluation of the neutron dose was done by adjusting the simulated photon dose to a common prostate delivery (e.g. 7560 cGy in 42 fractions) and finding the corresponding neutron dose yield from the simulation. A variance reduction technique was conducted for the neutrons yield and transported. Results: An effective dose of 3.65 cGy due to neutrons was found in the prostate volume. The dose to central areas of the prostate was found to be about 10 cGy. Conclusion: The neutron dose yielded does not justify a clinical implant of Beryllium seeds. Nevertheless, one should investigate the Neutron dose obtained when a larger Beryllium loading is combined with commercially available 40 MeV Linacs.

  3. Fluence and dose measurements for an accelerator neutron beam

    NASA Astrophysics Data System (ADS)

    Liu, Z.; Byun, S. H.; McNeill, F. E.; Mothersill, C. E.; Seymour, C. B.; Prestwich, W. V.

    2007-10-01

    The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 106 neutrons s-1). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.

  4. High-Yield D-T Neutron Generator

    SciTech Connect

    Ludewigt, B.A.; Wells, R.P.; Reijonen, J.

    2006-11-15

    A high-yield D-T neutron generator has been developed for neutron interrogation in homeland security applications such as cargo screening. The generator has been designed as a sealed tube with a performance goal of producing 5 {center_dot} 10{sup 11} n/s over a long lifetime. The key generator components developed are a radio-frequency (RF) driven ion source and a beam-loaded neutron production target that can handle a beam power of 10 kW. The ion source can provide a 100 mA D{sup +}/T{sup +} beam current with a high fraction of atomic species and can be pulsed up to frequencies of several kHz for pulsed neutron generator operation. Testing in D-D operation has been started.

  5. Calculation of dose contributions of electron and charged heavy particles inside phantoms irradiated by monoenergetic neutron.

    PubMed

    Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Yasushi; Miyahara, Nobuyuki

    2008-09-01

    The radiation-transport code PHITS with an event generator mode has been applied to analyze energy depositions of electrons and charged heavy particles in two spherical phantoms and a voxel-based mouse phantom upon neutron irradiation. The calculations using the spherical phantoms quantitatively clarified the type and energy of charged particles which are released through interactions of neutrons with the phantom elements and contribute to the radiation dose. The relative contribution of electrons increased with an increase in the size of the phantom and with a decrease in the energy of the incident neutrons. Calculations with the voxel-based mouse phantom for 2.0-MeV neutron irradiation revealed that the doses to different locations inside the body are uniform, and that the energy is mainly deposited by recoil protons. The present study has demonstrated that analysis using PHITS can yield dose distributions that are accurate enough for RBE evaluation. PMID:18580044

  6. Sensitivity of Measured Fission Yields on Prompt-neutron Corrections

    NASA Astrophysics Data System (ADS)

    Al-Adili, A.; Hambsch, F.-J.; Pomp, S.; Oberstedt, S.

    2014-05-01

    Although the number of emitted prompt neutrons from the fission fragments increases as a function of excitation energy, it is not fully understood whether the increase in νbar (A) as a function of En is mass dependent. The share of excitation energies among the fragments is still under debate, but there are reasons to believe that the excess in neutron emission originates only from the heavy fragments, leaving νbarlight (A) almost unchanged. We have investigated the consequences of a mass-dependent increase in νbar (A) on the final mass and energy distributions. The analysis have been performed on experimentally measured data on 234U (n, f). The assumptions concerning νbar (A) are essential when analysing measurements based on the 2E-technique, and impact significantly on the measured observables. For example, the post-neutron emission mass yield distribution revealed changes up to 10-30 %. The outcome of this work pinpoints the urgent need to determine νbar (A) experimentally, and in particular, how νbar (A) changes as a function of incident neutron energy. Many fission yields in the data libraries could be largely affected, since their analysis is based on a different assumption concerning the neutron emission.

  7. Estimation of neutron-induced spallation yields of krypton isotopes

    NASA Astrophysics Data System (ADS)

    Karol, Paul J.; Tobin, Michael J.; Shibata, Seiichi

    1983-10-01

    A procedure is outlined for estimating cross sections for neutron-induced spallation products relative to those for proton-induced reactions. When combined with known proton spallation systematics, it is demonstrated that cumulative yields for cosmogenically-important stable 84Kr and 86Kr isotopes are ~1.4 and ~2.8 times greater, respectively, for incident neutrons compared to protons at 0.2<=E<=3.0 GeV for nearby medium mass targets. Yields for lighter kryptons are relatively insensitive to the identity of the incident nucleon. NUCLEAR REACTIONS (n, spallation), 0.2<=En<=3.0 GeV, stable Kr product yield estimates from proton spallation systematics.

  8. Neutron emission and fragment yield in high-energy fission

    SciTech Connect

    Grudzevich, O. T. Klinov, D. A.

    2013-07-15

    The KRIS special library of spectra and emission probabilities in the decays of 1500 nuclei excited up to energies between 150 and 250 MeV was developed for correctly taking into account the decay of highly excited nuclei appearing as fission fragments. The emission of neutrons, protons, and photons was taken into account. Neutron emission fromprimary fragments was found to have a substantial effect on the formation of yields of postneutron nuclei. The library was tested by comparing the calculated and measured yields of products originating from the fission of nuclei that was induced by high-energy protons. The method for calculating these yields was tested on the basis of experimental data on the thermal-neutroninduced fission of {sup 235}U nuclei.

  9. Measurements of neutron dose rates with a balloon in Japan.

    PubMed

    Nagaoka, K; Hiraide, I; Sato, K; Yamagami, T; Nakamura, T; Yabutani, T

    2007-01-01

    Measurements of cosmic-ray neutron dose rates with a balloon in Sanriku, Japan (geographic location: 39 degrees N, 142 degrees E; corresponding geomagnetic latitude: 30 degrees N) were conducted at an altitude from 0.2 to 25 km on 25-26 August 2004 when solar activity was at an average level. Neutron dose rates given as ambient dose equivalent rates (H(10)) were measured with high-sensitive neutron dose equivalent counters and electronic silicon personal dosimeters (EPDs). The neutron dose rates increased with increasing altitude, but they were saturated around 15-20 km and decreased with increasing altitude beyond 20 km. The neutron ambient dose equivalent rate was 1.5 microSv/h(- 1) at 20 km. Measured values were corrected for the deviation of the energy response of the dose equivalent counter from the fluence-to-ambient dose equivalent conversion coefficient, and the corrected values were very close to the calculated values with EPCARD. On the other hand, neutron measurements by the EPDs gave about 10 times overestimation because of the high sensitivity to cosmic-ray protons.

  10. Characterization of neutron yield and x-ray spectra of a High Flux Neutron Generator (HFNG)

    NASA Astrophysics Data System (ADS)

    Nnamani, Nnaemeka; HFNG Collaboration

    2015-04-01

    The High Flux Neutron Generator (HFNG) is a DD plasma-based source, with a self-loading target intended for fundamental science and engineering applications, including 40 Ar/39 Ar geochronology, neutron cross section measurements, and radiation hardness testing of electronics. Our first estimate of the neutron yield, based on the population of the 4.486 hour 115 In isomer gave a neutron yield of the order 108 n/sec; optimization is ongoing to achieve the design target of 1011 n/sec. Preliminary x-ray spectra showed prominent energy peaks which are likely due to atomic line-emission from back-streaming electrons accelerated up to 100 keV impinging on various components of the HFNG chamber. Our x-ray and neutron diagnostics will aid us as we continue to evolve the design to suppress back-streaming electrons, necessary to achieve higher plasma beam currents, and thus higher neutron flux. This talk will focus on the characterization of the neutron yield and x-ray spectra during our tests. A collimation system is being installed near one of the chamber ports for improved observation of the x-ray spectra. This work is supported by NSF Grant No. EAR-0960138, U.S. DOE LBNL Contract No. DE-AC02-05CH11231, U.S. DOE LLNL Contract No. DE-AC52-07NA27344, and the UC Office of the President Award 12-LR-238745.

  11. Neutron yield saturation in plasma focus: A fundamental cause

    SciTech Connect

    Lee, S.

    2009-10-12

    Plasma focus research in the direction of fusion energy faces the limitation of observed neutron saturation; the neutron yield Y{sub n} falls away from Y{sub n}{approx}E{sub 0}{sup 2}, the scaling deteriorating as storage energy E{sub 0} increases toward 1 MJ. Numerical experiments confirm that Y{sub n}{approx}E{sub 0}{sup 2} applies at low energies and drops to Y{sub n}{approx}E{sub 0}{sup 0.8} toward 25 MJ; deteriorating already at several hundred kilojoules. We point out that the cause is the dynamic resistance of the axial phase that is constant for all plasma foci. This dynamic resistance dominates the circuit as capacitor bank surge impedance becomes insignificant at large E{sub 0}, causing current, hence neutron 'saturation'.

  12. Moderated 252Cf neutron energy spectra in brain tissue and calculated boron neutron capture dose.

    PubMed

    Rivard, Mark J; Zamenhof, Robert G

    2004-11-01

    While there is significant clinical experience using both low- and high-dose (252)Cf brachytherapy, combination therapy using (10)B for neutron capture therapy-enhanced (252)Cf brachytherapy has not been performed. Monte Carlo calculations were performed in a brain phantom (ICRU 44 brain tissue) to evaluate the dose enhancement predicted for a range of (10)B concentrations over a range of distances from a clinical (252)Cf source. These results were compared to experimental measurements and calculations published in the literature. For (10)B concentrations neutron capture dose enhancement was small in comparison to the (252)Cf fast neutron dose.

  13. A high yield neutron target for cancer therapy

    NASA Technical Reports Server (NTRS)

    Alger, D. L.; Steinberg, R.

    1972-01-01

    A rotating target was developed that has the potential for providing an initial yield of 10 to the 13th power neutrons per second by the T(d,n)He-4 reaction, and a useable lifetime in excess of 600 hours. This yield and lifetime are indicated for a 300 Kv and 30 mA deuteron accelerator and a 30 microns thick titanium tritide film formed of the stoichiometric compound TiT2. The potential for extended lifetime is made possible by incorporating a sputtering electrode that permits use of titanium tritide thicknesses much greater than the deuteron range. The electrode is used to remove in situ depleted titanium layers to expose fresh tritide beneath. The utilization of the rotating target as a source of fast neutrons for cancer therapy is discussed.

  14. Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

    NASA Astrophysics Data System (ADS)

    Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2014-09-01

    A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

  15. Measurements of DT and DD neutron yields by neutron activation on TFTR

    SciTech Connect

    Barnes, C.W.; Larson, A.R.; LeMunyan, G.; Loughlin, M.J.

    1994-05-05

    A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10{sup 12} to over 10{sup 18}, and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants. and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the {plus_minus}9% (one-sigma,) accuracy of the measurements: also agreeing are yields from silicon foils using the ACTL library cross-section. While the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the {sup 115}In(n,n) {sup 115m}In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments.

  16. Neutron source capability assessment for cumulative fission yields measurements

    SciTech Connect

    Descalle, M A; Dekin, W; Kenneally, J

    2011-04-06

    A recent analysis of high-quality cumulative fission yields data for Pu-239 published in the peer-reviewed literature showed that the quoted experimental uncertainties do not allow a clear statement on how the fission yields vary as a function of energy. [Prussin2009] To make such a statement requires a set of experiments with well 'controlled' and understood sources of experimental errors to reduce uncertainties as low as possible, ideally in the 1 to 2% range. The Inter Laboratory Working Group (ILWOG) determined that Directed Stockpile Work (DSW) would benefit from an experimental program with the stated goal to reduce the measurement uncertainties significantly in order to make a definitive statement of the relationship of energy dependence to the cumulative fission yields. Following recent discussions between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), there is a renewed interest in developing a concerted experimental program to measure fission yields in a neutron energy range from thermal energy (0.025 eV) to 14 MeV with an emphasis on discrete energies from 0.5 to 4 MeV. Ideally, fission yields would be measured at single energies, however, in practice there are only 'quasi-monoenergetic' neutrons sources of finite width. This report outlines a capability assessment as of June 2011 of available neutron sources that could be used as part of a concerted experimental program to measure cumulative fission yields. In a framework of international collaborations, capabilities available in the United States, at the Atomic Weapons Establishment (AWE) in the United Kingdom and at the Commissariat Energie Atomique (CEA) in France are listed. There is a need to develop an experimental program that will reduce the measurement uncertainties significantly in order to make a definitive statement of the relationship of energy dependence to the cumulative fission yields. Fission and monoenergetic neutron sources are available that

  17. Measurement of Fission Product Yields from Fast-Neutron Fission

    NASA Astrophysics Data System (ADS)

    Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  18. Evaluation of absorbed dose in Gadolinium neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Abdullaeva, Gayane; Djuraeva, Gulnara; Kim, Andrey; Koblik, Yuriy; Kulabdullaev, Gairatulla; Rakhmonov, Turdimukhammad; Saytjanov, Shavkat

    2015-02-01

    Gadolinium neutron capture therapy (GdNCT) is used for treatment of radioresistant malignant tumors. The absorbed dose in GdNCT can be divided into four primary dose components: thermal neutron, fast neutron, photon and natural gadolinium doses. The most significant is the dose created by natural gadolinium. The amount of gadolinium at the irradiated region is changeable and depends on the gadolinium delivery agent and on the structure of the location where the agent is injected. To de- fine the time dependence of the gadolinium concentration ρ(t) in the irradiated region the pharmacokinetics of gadolinium delivery agent (Magnevist) was studied at intratumoral injection in mice and intramuscular injection in rats. A polynomial approximation was applied to the experimental data and the influence of ρ(t) on the relative change of the absorbed dose of gadolinium was studied.

  19. Neutron-induced gamma dose from a reactor beam filter for boron neutron capture therapy.

    PubMed

    Harrington, B V

    1989-01-01

    For the boron neutron capture therapy (NCT) of deep-seated metastatic melanoma, an epithermal (up to a few keV energy) neutron beam from a reactor horizontal facility could be useful if the inherent contamination from fast neutrons and gamma rays could be minimised. Calculations for ANSTO's 10 MW research reactor HIFAR have shown that, even though a filter material such as AlF3 attenuates the fast neutron dose, the beam quality improvement is counteracted by a relative increase in the gamma dose because of the gammas arising from neutron captures in the filter material, particularly the aluminium. The aluminium gammas, most of which arise from thermal neutron capture, are hard and cannot be attenuated by lead or bismuth without comparable attenuation of the epithermal neutron flux. Addition of an absorber such as 6Li to the AlF3 filter was investigated as a means of reducing the hard gamma dose, but the improvement in beam quality was small and at considerable cost to dose intensity. Dose characteristics calculations confirmed the superiority of a tangential beam over a radial beam with better results from an unfiltered tangential beam than from an AlF3 filter in a radial beam. This study showed conclusively that assessments of filter assemblies based on the effect of individual components on either the neutron or gamma dose in isolation are inadequate. In assessing any epithermal neutron filter, thermal neutron shield, and gamma shield combination, the total effect of each on the neutron, gamma, and boron-10 dose must be considered.

  20. Low doses of neutrons induce changes in gene expression

    SciTech Connect

    Woloschak, G.E.; Chang-Liu, C.M. ); Panozzo, J.; Libertin, C.R. )

    1993-01-01

    Studies were designed to identify genes induced following low-dose neutron but not following [gamma]-ray exposure in fibroblasts. Our past work had shown differences in the expression of [beta]-protein kinase C and c-fos genes, both being induced following [gamma]-ray but not neutron exposure. We have identified two genes that are induced following neutron, but not [gamma]-ray, exposure: Rp-8 (a gene induced by apoptosis) and the long terminal repeat (LTR) of the human immunodeficiency (HIV). Rp-8 mRNA induction was demonstrated in Syrian hamster embryo fibroblasts and was found to be induced in cells exposed to neutrons administered at low (0.5 cGy/min) and at high dose rate (12 cGy/min). The induction of transcription from the LTR of HIV was demonstrated in HeLa cells bearing a transfected construct of the chloramphenicol acetyl transferase (CAT) gene driven by the HIV-LTR promoter. Measures of CAT activity and CAT transcripts following irradiation demonstrated an unresponsiveness to [gamma] rays over a broad range of doses. Twofold induction of the HIV-LTR was detected following neutron exposure (48 cGy) administered at low (0.5 cGy/min) but not high (12 cGy/min) dose rates. Ultraviolet-mediated HIV-LTR induction was inhibited by low-dose-rate neutron exposure.

  1. Low doses of neutrons induce changes in gene expression

    SciTech Connect

    Woloschak, G.E.; Chang-Liu, C.M.; Panozzo, J.; Libertin, C.R.

    1993-06-01

    Studies were designed to identify genes induced following low-dose neutron but not following {gamma}-ray exposure in fibroblasts. Our past work had shown differences in the expression of {beta}-protein kinase C and c-fos genes, both being induced following {gamma}-ray but not neutron exposure. We have identified two genes that are induced following neutron, but not {gamma}-ray, exposure: Rp-8 (a gene induced by apoptosis) and the long terminal repeat (LTR) of the human immunodeficiency (HIV). Rp-8 mRNA induction was demonstrated in Syrian hamster embryo fibroblasts and was found to be induced in cells exposed to neutrons administered at low (0.5 cGy/min) and at high dose rate (12 cGy/min). The induction of transcription from the LTR of HIV was demonstrated in HeLa cells bearing a transfected construct of the chloramphenicol acetyl transferase (CAT) gene driven by the HIV-LTR promoter. Measures of CAT activity and CAT transcripts following irradiation demonstrated an unresponsiveness to {gamma} rays over a broad range of doses. Twofold induction of the HIV-LTR was detected following neutron exposure (48 cGy) administered at low (0.5 cGy/min) but not high (12 cGy/min) dose rates. Ultraviolet-mediated HIV-LTR induction was inhibited by low-dose-rate neutron exposure.

  2. Neutron detector simultaneously measures fluence and dose equivalent

    NASA Technical Reports Server (NTRS)

    Dvorak, R. F.; Dyer, N. C.

    1967-01-01

    Neutron detector acts as both an area monitoring instrument and a criticality dosimeter by simultaneously measuring dose equivalent and fluence. The fluence is determined by activation of six foils one inch below the surface of the moderator. Dose equivalent is determined from activation of three interlocked foils at the center of the moderator.

  3. Estimation of Secondary Neutron Dose during Proton Therapy

    NASA Astrophysics Data System (ADS)

    Urban, Tomas; Klusoň, Jaroslav

    2014-06-01

    During proton radiotherapy, secondary neutrons are produced by nuclear interactions in the material along the beam path, in the treatment nozzle (including the fixed scatterer, range modulator, etc.) and, of course, after entering the patient. The dose equivalent deposited by these neutrons is usually not considered in routine treatment planning. In this study, there has been estimated the neutron dose in patient (in as well as around the target volume) during proton radiotherapy using scattering and scanning techniques. The proton induced neutrons (and photons) have been simulated in the simple geometry of the single scattering and the pencil beam scanning universal nozzles and in geometry of the plastic phantom (made of tissue equivalent material - RW3 - imitate the patient). In simulations of the scattering nozzle, different types of brass collimators have been used as well. Calculated data have been used as an approximation of the radiation field in and around the chosen/potential target volume in the patient (plastic phantom). For the dose equivalent evaluation, fluence-to-dose conversion factors from ICRP report have been employed. The results of calculated dose from neutrons in various distances from the spot for different treatment technique and for different energies of incident protons have been compared and evaluated in the context of the dose deposited in the target volume. This work was supported by RVO: 68407700 and Grant Agency of the CTU in Prague, grant No. SGS12/200/OHK4/3T/14.

  4. The heavy element yields of neutron capture nucleosynthesis

    NASA Technical Reports Server (NTRS)

    Cameron, A. G. W.

    1982-01-01

    Consideration of the contribution made to the abundances of the heavy element isotopes by the S- and R-processes of nucleosynthesis has led to the determination that the previous assumption concerning the exclusive alignment of isobars to one or the other of these processes is probably in error. If the relatively small odd and even mass number abundance fluctuations characterizing R-process abundances are always the case, as assumed by this study, S-process contributions to the abundances of R-process isobars are substantial, consistent with transient flashing episodes in the S-process neutron production processes. A smooth and monotonically-decreasing curve of the abundance of the S-process yields times the neutron capture cross-section versus mass number is therefore the primary tool for the separation of the abundances due to the two processes.

  5. Neutron dose estimation in a zero power nuclear reactor

    NASA Astrophysics Data System (ADS)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  6. Characterization of a Pulse Neutron Source Yield under Field Conditions

    SciTech Connect

    Barzilov, Alexander; Novikov, Ivan; Womble, Phillip C.; Hopper, Lindsay

    2009-03-10

    Technique of rapid evaluation of a pulse neutron sources such as neutron generators under field conditions has been developed. The phoswich sensor and pulse-shape discrimination techniques have been used for the simultaneous measurements of fast neutrons, thermal neutrons, and photons. The sensor has been calibrated using activation neutron detectors and a pulse deuterium-tritium fusion neutron source.

  7. Monte Carlo simulation of the neutron monitor yield function

    NASA Astrophysics Data System (ADS)

    Mangeard, P.-S.; Ruffolo, D.; Sáiz, A.; Madlee, S.; Nutaro, T.

    2016-08-01

    Neutron monitors (NMs) are ground-based detectors that measure variations of the Galactic cosmic ray flux at GV range rigidities. Differences in configuration, electronics, surroundings, and location induce systematic effects on the calculation of the yield functions of NMs worldwide. Different estimates of NM yield functions can differ by a factor of 2 or more. In this work, we present new Monte Carlo simulations to calculate NM yield functions and perform an absolute (not relative) comparison with the count rate of the Princess Sirindhorn Neutron Monitor (PSNM) at Doi Inthanon, Thailand, both for the entire monitor and for individual counter tubes. We model the atmosphere using profiles from the Global Data Assimilation System database and the Naval Research Laboratory Mass Spectrometer, Incoherent Scatter Radar Extended model. Using FLUKA software and the detailed geometry of PSNM, we calculated the PSNM yield functions for protons and alpha particles. An agreement better than 9% was achieved between the PSNM observations and the simulated count rate during the solar minimum of December 2009. The systematic effect from the electronic dead time was studied as a function of primary cosmic ray rigidity at the top of the atmosphere up to 1 TV. We show that the effect is not negligible and can reach 35% at high rigidity for a dead time >1 ms. We analyzed the response function of each counter tube at PSNM using its actual dead time, and we provide normalization coefficients between count rates for various tube configurations in the standard NM64 design that are valid to within ˜1% for such stations worldwide.

  8. An angular dependent neutron effective-dose-equivalent dosimeter

    NASA Astrophysics Data System (ADS)

    Veinot, Kenneth Guy

    The effective-dose-equivalent (EDE) is a strong function of angular orientation in a radiation field. Detection systems that attempt to measure the EDE directly would be desirable. Historically, dosimeters have been designed to respond as isotropically as possible in a radiation field. However, since the EDE is strongly dependent upon the incident angle of the radiation, past designs are no longer desirable for personal radiation dosimetry. In addition, the EDE is a function of incident neutron energy. CR-39 foils are commonly used neutron detectors. Neutrons produce tracks in CR-39 (allyl diglycol polycarbonate) detectors over a wide energy range. Through chemical or electrochemical etching, these tracks can be enlarged and counted. From this track count, the fluence of neutrons incident on the CR-39 foils may be inferred. Thermoluminescent dosimeters (TLDs) are another method of neutron detection. Both of these detectors have angular response properties. In the present work, calculations of EDE were compared to calculations and measurements of the angular responses of CR-39 and TLD neutron dosimeters. The measurements used a variety of neutron sources, each with its own characteristic energy spectrum. This research resulted in a neutron personal dosimeter prototype whose angular response properties resembled the angular response of EDE.

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

    SciTech Connect

    Brackenbush, L.W.; Reece, W.D.; Miller, S.D.; Endres, G.W.R.; Durham, J.S.; Scherpelz, R.I.; Tomeraasen, P.L.; Stroud, C.M.; Faust, L.G.; Vallario, E.J.

    1988-07-01

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

  10. Neutron absorbed dose determination by calculations of recoil energy.

    PubMed

    Wrobel, F; Benabdesselam, M; Iacconi, P; Lapraz, D

    2004-01-01

    The aim of this work is to calculate the absorbed dose to matter due to neutrons in the 5-150 MeV energy range. Materials involved in the calculations are Al2O3, CaSO4 and CaS, which may be used as dosemeters and have already been studied for their luminescent properties. The absorbed dose is assumed to be mainly due to the energy deposited by the recoils. Elastic reactions are treated with the ECIS code while for the non-elastic ones, a Monte Carlo code has been developed and allowed to follow the nucleus decay and to determine its characteristics (nature and energy). Finally, the calculations show that the absorbed dose is mainly due to non-elastic process and that above 20 MeV this dose decreases slightly with the neutron energy. PMID:15353750

  11. Measurement of neutron energy spectra and neutron dose rates from 7Li(p,n)7Be reaction induced on thin LiF target

    NASA Astrophysics Data System (ADS)

    Atanackovic, Jovica; Matysiak, Witold; Dubeau, Jacques; Witharana, Sampath; Waker, Anthony

    2015-02-01

    The measurements of neutron energy spectra and neutron dose rates were performed using the KN Van de Graaff accelerator, located at the McMaster University Accelerator Laboratory (MAL). Protons were accelerated on the thin lithium fluoride (LiF) target and produced mono-energetic neutrons which were measured using three different spectrometers: Bonner Sphere Spectrometer (BSS), Nested Neutron Spectrometer (NNS), and Rotational Proton Recoil Spectrometer (ROSPEC). The purpose of this work is (1) measurement and quantification of low energy accelerator neutron fields in terms of neutron fluence and dose, (2) comparison of results obtained by three different instruments, (3) comparison of measurements with Monte Carlo simulations based on theoretical neutron yields from 7Li(p,n)7Be nuclear reaction, and (4) comparison of results obtained using different neutron spectral unfolding methods. The nominal thickness of the LiF target used in the experiment was 50 μg /cm2, which corresponds to the linear thickness of 0.19 μm and results in approximately 6 keV energy loss for the proton energies used in the experiment (2.2, 2.3, 2.4 and 2.5 MeV). For each of the proton energies, neutron fluence per incident proton charge was measured and several dosimetric quantities of interest in radiation protection were derived. In addition, theoretical neutron yield calculations together with the results of Monte Carlo (MCNP) modeling of the neutron spectra are reported. Consistent neutron fluence spectra were obtained with three detectors and good agreement was observed between theoretically calculated and measured neutron fluences and derived dosimetric quantities for investigated proton energies at 2.3, 2.4 and 2.5 MeV. In the case of 2.2 MeV, some plausibly explainable discrepancies were observed.

  12. Measurement of neutron spectra generated from bombardment of 4 to 24 MeV protons on a thick 9Be target and estimation of neutron yields

    NASA Astrophysics Data System (ADS)

    Paul, Sabyasachi; Sahoo, G. S.; Tripathy, S. P.; Sharma, S. C.; Ramjilal, Ninawe, N. G.; Sunil, C.; Gupta, A. K.; Bandyopadhyay, T.

    2014-06-01

    A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK_n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed.

  13. Measurement of neutron spectra generated from bombardment of 4 to 24 MeV protons on a thick ⁹Be target and estimation of neutron yields.

    PubMed

    Paul, Sabyasachi; Sahoo, G S; Tripathy, S P; Sharma, S C; Ramjilal; Ninawe, N G; Sunil, C; Gupta, A K; Bandyopadhyay, T

    2014-06-01

    A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK_n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed. PMID:24985813

  14. Measurement of neutron spectra generated from bombardment of 4 to 24 MeV protons on a thick {sup 9}Be target and estimation of neutron yields

    SciTech Connect

    Paul, Sabyasachi; Sahoo, G. S.; Tripathy, S. P. E-mail: tripathy@barc.gov.in; Sunil, C.; Bandyopadhyay, T.; Sharma, S. C.; Ramjilal,; Ninawe, N. G.; Gupta, A. K.

    2014-06-15

    A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK-n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed.

  15. Picosecond Neutron Yields from Ultra-Intense Laser-Target Interactions

    NASA Astrophysics Data System (ADS)

    Ellison, C. Leland; Fuchs, Julien

    2009-11-01

    High-flux neutron sources for neutron imaging and materials analysis applications have typically been provided by accelerator-based (Spallation Neutron Source) and reactor-based (High Flux Isotope Reactor) neutron sources. A novel approach is to use ultra-intense (> 10^18 W/cm^2) laser-target interactions to generate picosecond, collimated neutrons. Here we examine the feasibility of a source based on current (LULI) and upcoming laser facility capabilities. A Monte-Carlo code calculates angular and energy distributions of neutrons generated by D-D fusion events occurring within a deuterated target for a given incident beam of D+ ions. The parameters of the deuteron beam are well understood from laser-plasma and laser-target studies relevant to fast-ignition fusion. Expected neutron yields are presented in comparison to conventional neutron sources, previous experimental neutron yields, and within the context of neutron shielding safety requirements.

  16. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    NASA Astrophysics Data System (ADS)

    Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

    2013-11-01

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18O(p, n)18F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H218O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection.

  17. New calculations of neutron kerma coefficients and dose equivalent.

    PubMed

    Liu, Zhenzhou; Chen, Jinxiang

    2008-06-01

    For neutron energies ranging from 1 keV to 20 MeV, the kerma coefficients for elements H, C, N, O, light water, and ICRU tissue were deduced respectively from microscopic cross sections and Monte Carlo simulation (MCNP code). The results are consistent within admitted uncertainties with values evaluated by an international group (Chadwick et al 1999 Med. Phys. 26 974-91). The ambient dose equivalent generated in the ISO-recommended neutron field for an Am-Be neutron source (ISO 8529-1: 2001(E)) was obtained from the kerma coefficients and Monte Carlo calculation. In addition, it was calculated directly by multiplying the neutron fluence by the fluence-to-ambient dose conversion coefficients recommended by ICRP (ICRP 1996 ICRP Publication 74 (Oxford: Pergamon)). The two results agree well with each other. The main feature of this work is our Monte Carlo simulation design and the treatments differing from the work of others in the calculation of neutron energy transfer in non-elastic processes. PMID:18495982

  18. Cation disorder in high-dose, neutron-irradiated spinel

    SciTech Connect

    Sickafus, K.E.; Larson, A.C.; Yu, N.

    1995-04-01

    The objective of this effort is to determine whether MgAl{sub 2}O{sub 4} spinel is a suitable ceramic for fusion applications. The crystal structures of MgAl{sub 2}O{sub 4} spinel single crystals irradiated to high neutron fluences [>5{times}10{sup 26} n/m{sup 2} (E{sub n}>0.1 MeV)] were examined by neutron diffraction. Crystal structure refinement of the highese dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by {approx}20% while increasing by {approx}8% on octahedral sites.

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

    PubMed

    Tsujimura, N; Yoshida, T; Takada, C

    2011-07-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. PMID:21498409

  20. Analysis for Radiation and Shielding Dose in Plasma Focus Neutron Source Using FLUKA

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

    Monte Carlo simulations have been performed for the attenuation of neutron radiation produced at Plasma focus (PF) devices through various shielding design. At the test site it will be fired with deuterium and tritium (D-T) fusion resulting in a yield of about 1013 fusion neutrons of 14 MeV. This poses a radiological hazard to scientists and personnel operating the device. The goal of this paper was to evaluate various shielding options under consideration for the PF operating with D-T fusion. Shields of varying neutrons-shielding effectiveness were investigated using concrete, polyethylene, paraffin and borated materials. The most effective shield, a labyrinth structure, allowed almost 1,176 shots per year while keeping personnel under 20 mSV of dose. The most expensive shield that used, square shield with 100 cm concrete thickness on the walls and Borated paraffin along with borated polyethylene added outside the concrete allowed almost 15,000 shot per year.

  1. Measuring the absolute DT neutron yield using the Magnetic Recoil Spectrometer at OMEGA and the NIF

    SciTech Connect

    Mackinnon, A; Casey, D; Frenje, J A; Johnson, M G; Seguin, F H; Li, C K; Petrasso, R D; Glebov, V Y; Katz, J; Knauer, J; Meyerhofer, D; Sangster, T; Bionta, R; Bleuel, D; Hachett, S P; Hartouni, E; Lepape, S; Mckernan, M; Moran, M; Yeamans, C

    2012-05-03

    A Magnetic Recoil Spectrometer (MRS) has been installed and extensively used on OMEGA and the National Ignition Facility (NIF) for measurements of the absolute neutron spectrum from inertial confinement fusion (ICF) implosions. From the neutron spectrum measured with the MRS, many critical implosion parameters are determined including the primary DT neutron yield, the ion temperature, and the down-scattered neutron yield. As the MRS detection efficiency is determined from first principles, the absolute DT neutron yield is obtained without cross-calibration to other techniques. The MRS primary DT neutron measurements at OMEGA and the NIF are shown to be in excellent agreement with previously established yield diagnostics on OMEGA, and with the newly commissioned nuclear activation diagnostics on the NIF.

  2. Personnel neutron dose assessment upgrade: Volume 1, Personnel neutron dosimetry assessment: (Final report)

    SciTech Connect

    Hadlock, D.E.; Brackenbush, L.W.; Griffith, R.V.; Hankins, D.E.; Parkhurst, M.A.; Stroud, C.M.; Faust, L.G.; Vallario, E.J.

    1988-07-01

    This report provides guidance on the characteristics, use, and calibration criteria for personnel neutron dosimeters. The report is applicable for neutrons with energies ranging from thermal to less than 20 MeV. Background for general neutron dosimetry requirements is provided, as is relevant federal regulations and other standards. The characteristics of personnel neutron dosimeters are discussed, with particular attention paid to passive neutron dosimetry systems. Two of the systems discussed are used at DOE and DOE-contractor facilities (nuclear track emulsion and thermoluminescent-albedo) and another (the combination TLD/TED) was recently developed. Topics discussed in the field applications of these dosimeters include their theory of operation, their processing, readout, and interpretation, and their advantages and disadvantages for field use. The procedures required for occupational neutron dosimetry are discussed, including radiation monitoring and the wearing of dosimeters, their exchange periods, dose equivalent evaluations, and the documenting of neutron exposures. The coverage of dosimeter testing, maintenance, and calibration includes guidance on the selection of calibration sources, the effects of irradiation geometries, lower limits of detectability, fading, frequency of calibration, spectrometry, and quality control. 49 refs., 6 figs., 8 tabs.

  3. SECONDARY NEUTRON DOSES IN A PROTON THERAPY CENTRE.

    PubMed

    De Saint-Hubert, M; Saldarriaga Vargas, C; Van Hoey, O; Schoonjans, W; De Smet, V; Mathot, G; Stichelbaut, F; Manessi, G; Dinar, N; Aza, E; Cassell, C; Silari, M; Vanhavere, F

    2016-09-01

    The formation of secondary high-energy neutrons in proton therapy can be a concern for radiation protection of staff. In this joint intercomparative study (CERN, SCK•CEN and IBA/IRISIB/ULB), secondary neutron doses were assessed with different detectors in several positions in the Proton Therapy Centre, Essen (Germany). The ambient dose equivalent H(*)(10) was assessed with Berthold LB 6411, WENDI-2, tissue-equivalent proportional counter (TEPC) and Bonner spheres (BS). The personal dose equivalent Hp(10) was measured with two types of active detectors and with bubble detectors. Using spectral and basic angular information, the reference Hp(10) was estimated. Results concerning staff exposure show H(*)(10) doses between 0.5 and 1 nSv/monitoring unit in a technical room. The LB 6411 showed an underestimation of H(*)(10), while WENDI-2 and TEPC showed good agreement with the BS data. A large overestimation for Hp(10) was observed for the active personal dosemeters, while the bubble detectors showed only a slight overestimation.

  4. Dose inhomogeneities for photons and neutrons near interfaces.

    PubMed

    Broerse, J J; Zoetelief, J

    2004-01-01

    Perturbations of charged particle equilibrium (CPE) at interfaces of materials of different atomic composition can lead to considerable differences in the energy deposition by photons and neutrons. Specific examples of these interface perturbations are encountered during irradiation of body cavities and soft tissue adjacent to bone or metallic implants and irradiation of cells in monolayer on the bottom of culture dishes. Another example is the build-up of CPE at air-tissue interfaces, referred to in radiotherapy as the skin sparing effect. For photon irradiation excess production of secondary electrons in high-Z materials, such as glass, bone or gold, will induce appreciably higher doses and decreased cell survival compared to the equilibrium situation. The energy dissipation of fast neutrons in biological materials occurs through recoil protons, heavy recoil nuclei and products of nuclear reactions. Owing to the large contribution from recoil protons to the neutron kerma, the hydrogen content of the biological material mainly determines the energy deposition. For neutron irradiation of cells in monolayer, CPE can be established or deliberately avoided by mounting tissue-equivalent plastic or carbon discs in front of the cells, respectively. This approach makes it possible to distinguish the biological effects of the low- and high-LET radiation components. PMID:15623886

  5. Analysis of incident-energy dependence of delayed neutron yields in actinides

    SciTech Connect

    Nasir, Mohamad Nasrun bin Mohd Metorima, Kouhei Ohsawa, Takaaki Hashimoto, Kengo

    2015-04-29

    The changes of delayed neutron yields (ν{sub d}) of Actinides have been analyzed for incident energy up to 20MeV using realized data of precursor after prompt neutron emission, from semi-empirical model, and delayed neutron emission probability data (P{sub n}) to carry out a summation method. The evaluated nuclear data of the delayed neutron yields of actinide nuclides are still uncertain at the present and the cause of the energy dependence has not been fully understood. In this study, the fission yields of precursor were calculated considering the change of the fission fragment mass yield based on the superposition of fives Gaussian distribution; and the change of the prompt neutrons number associated with the incident energy dependence. Thus, the incident energy dependent behavior of delayed neutron was analyzed.The total number of delayed neutron is expressed as ν{sub d}=∑Y{sub i} • P{sub ni} in the summation method, where Y{sub i} is the mass yields of precursor i and P{sub ni} is the delayed neutron emission probability of precursor i. The value of Y{sub i} is derived from calculation of post neutron emission mass distribution using 5 Gaussian equations with the consideration of large distribution of the fission fragments. The prompt neutron emission ν{sub p} increases at higher incident-energy but there are two different models; one model says that the fission fragment mass dependence that prompt neutron emission increases uniformly regardless of the fission fragments mass; and the other says that the major increases occur at heavy fission fragments area. In this study, the changes of delayed neutron yields by the two models have been investigated.

  6. Neutron yield enhancement in laser-induced deuterium-deuterium fusion using a novel shaped target.

    PubMed

    Zhao, J R; Zhang, X P; Yuan, D W; Chen, L M; Li, Y T; Fu, C B; Rhee, Y J; Li, F; Zhu, B J; Li, Yan F; Liao, G Q; Zhang, K; Han, B; Liu, C; Huang, K; Ma, Y; Li, Yi F; Xiong, J; Huang, X G; Fu, S Z; Zhu, J Q; Zhao, G; Zhang, J

    2015-06-01

    Neutron yields have direct correlation with the energy of incident deuterons in experiments of laser deuterated target interaction [Roth et al., Phys. Rev. Lett. 110, 044802 (2013) and Higginson et al., Phys. Plasmas 18, 100703 (2011)], while deuterated plasma density is also an important parameter. Experiments at the Shenguang II laser facility have produced neutrons with energy of 2.45 MeV using d (d, n) He reaction. Deuterated foil target and K-shaped target were employed to study the influence of plasma density on neutron yields. Neutron yield generated by K-shaped target (nearly 10(6)) was two times higher than by foil target because the K-shaped target results in higher density plasma. Interferometry and multi hydro-dynamics simulation confirmed the importance of plasma density for enhancement of neutron yields.

  7. Neutron yield enhancement in laser-induced deuterium-deuterium fusion using a novel shaped target

    SciTech Connect

    Zhao, J. R.; Chen, L. M. Li, Y. T.; Li, F.; Zhu, B. J.; Li, Yan. F.; Liao, G. Q.; Huang, K.; Ma, Y.; Li, Yi. F.; Zhang, X. P.; Fu, C. B.; Yuan, D. W.; Zhang, K.; Han, B.; Zhao, G.; Rhee, Y. J.; Liu, C.; Xiong, J.; Huang, X. G.; and others

    2015-06-15

    Neutron yields have direct correlation with the energy of incident deuterons in experiments of laser deuterated target interaction [Roth et al., Phys. Rev. Lett. 110, 044802 (2013) and Higginson et al., Phys. Plasmas 18, 100703 (2011)], while deuterated plasma density is also an important parameter. Experiments at the Shenguang II laser facility have produced neutrons with energy of 2.45 MeV using d (d, n) He reaction. Deuterated foil target and K-shaped target were employed to study the influence of plasma density on neutron yields. Neutron yield generated by K-shaped target (nearly 10{sup 6}) was two times higher than by foil target because the K-shaped target results in higher density plasma. Interferometry and multi hydro-dynamics simulation confirmed the importance of plasma density for enhancement of neutron yields.

  8. A bounding estimate of neutron dose based on measured photon dose around single pass reactors at the Hanford site.

    PubMed

    Taulbee, Timothy D; Glover, Samuel E; Macievic, Gregory V; Hunacek, Mickey; Smith, Cheryl; DeBord, Gary W; Morris, Donald; Fix, Jack

    2010-07-01

    Neutron and photon radiation survey records have been used to evaluate and develop a neutron to photon (NP) ratio to reconstruct neutron doses to workers around Hanford's single pass reactors that operated from 1945 to 1972. A total of 5,773 paired neutron and photon measurements extracted from 57 boxes of survey records were used in the development of the NP ratio. The development of the NP ratio enables the use of the recorded dose from an individual's photon dosimeter badge to be used to estimate the unmonitored neutron dose. The Pearson rank correlation between the neutron and photon measurements was 0.71. The NP ratio best fit a lognormal distribution with a geometric mean (GM) of 0.8, a geometric standard deviation (GSD) of 2.95, and the upper 95 th % of this distribution was 4.75. An estimate of the neutron dose based on this NP ratio is considered bounding due to evidence that up to 70% of the total photon exposure received by workers around the single pass reactors occurs during shutdown maintenance and refueling activities when there is no significant neutron exposure. Thus when this NP ratio is applied to the total measured photon dose from an individual film badge dosimeter, the resulting neutron dose is considered bounded.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

  10. Deuterium-tritium neutron yield measurements with the 4.5 m neutron-time-of-flight detectors at NIF.

    PubMed

    Moran, M J; Bond, E J; Clancy, T J; Eckart, M J; Khater, H Y; Glebov, V Yu

    2012-10-01

    The first several campaigns of laser fusion experiments at the National Ignition Facility (NIF) included a family of high-sensitivity scintillator∕photodetector neutron-time-of-flight (nTOF) detectors for measuring deuterium-deuterium (DD) and DT neutron yields. The detectors provided consistent neutron yield (Y(n)) measurements from below 10(9) (DD) to nearly 10(15) (DT). The detectors initially demonstrated detector-to-detector Y(n) precisions better than 5%, but lacked in situ absolute calibrations. Recent experiments at NIF now have provided in situ DT yield calibration data that establish the absolute sensitivity of the 4.5 m differential tissue harmonic imaging (DTHI) detector with an accuracy of ± 10% and precision of ± 1%. The 4.5 m nTOF calibration measurements also have helped to establish improved detector impulse response functions and data analysis methods, which have contributed to improving the accuracy of the Y(n) measurements. These advances have also helped to extend the usefulness of nTOF measurements of ion temperature and downscattered neutron ratio (neutron yield 10-12 MeV divided by yield 13-15 MeV) with other nTOF detectors.

  11. Deuterium-tritium neutron yield measurements with the 4.5 m neutron-time-of-flight detectors at NIF

    SciTech Connect

    Moran, M. J.; Bond, E. J.; Clancy, T. J.; Eckart, M. J.; Khater, H. Y.; Glebov, V. Yu.

    2012-10-15

    The first several campaigns of laser fusion experiments at the National Ignition Facility (NIF) included a family of high-sensitivity scintillator/photodetector neutron-time-of-flight (nTOF) detectors for measuring deuterium-deuterium (DD) and DT neutron yields. The detectors provided consistent neutron yield (Y{sub n}) measurements from below 10{sup 9} (DD) to nearly 10{sup 15} (DT). The detectors initially demonstrated detector-to-detector Y{sub n} precisions better than 5%, but lacked in situ absolute calibrations. Recent experiments at NIF now have provided in situ DT yield calibration data that establish the absolute sensitivity of the 4.5 m differential tissue harmonic imaging (DTHI) detector with an accuracy of {+-}10% and precision of {+-}1%. The 4.5 m nTOF calibration measurements also have helped to establish improved detector impulse response functions and data analysis methods, which have contributed to improving the accuracy of the Y{sub n} measurements. These advances have also helped to extend the usefulness of nTOF measurements of ion temperature and downscattered neutron ratio (neutron yield 10-12 MeV divided by yield 13-15 MeV) with other nTOF detectors.

  12. Cation disorder in high dose neutron irradiated spinel

    SciTech Connect

    Sickafus, K.E.; Larson, A.C.; Yu, N.; Nastasi, M.; Hollenberg, G.W.; Garner, F.A.; Bradt, R.C.

    1994-06-01

    The crystal structures of MgAl{sub 2}O{sub 4} spinel single crystals irradiated to high neutron fluences (>5{center_dot}10{sup 26} n/m{sup 2} (E{sub n}>0.1 MeV)), were examined by neutron diffraction. Crystal structure refinement of the highest dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by {approximately}20% while increasing by {approximately}8% on octahedral sites. Since the neutron scattering length for Mg is considerably larger than for Al, this result is consistent with site exchange between Mg{sup 2+} ions on tetrahedral sites and Al{sup 3+} ions on octahedral sites. Least squares refinements also indicated that in all irradiated samples, at least 35% of Mg{sup 2+} and Al{sup 3+} ions in the crystal experienced disordering replacements. This retained dpa on the cation sublattices is the largest retained damage ever measured in an irradiated spinel material.

  13. Prediction of In-Phantom Dose Distribution Using In-Air Neutron Beam Characteristics for Boron Neutron Capture Synovectomy

    SciTech Connect

    Verbeke, Jerome M.; Chen, Allen S.; Vujic, Jasmina L.; Leung, Ka-Ngo

    2000-08-15

    A monoenergetic neutron beam simulation study was carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints such as knees and fingers. This study focuses on human knee joints. Two figures of merit are used to measure the neutron beam quality, the ratio of the synovium-absorbed dose to the skin-absorbed dose, and the ratio of the synovium-absorbed dose to the bone-absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment and that (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce the particle transport simulation time by a factor of 10 by modeling the moderator only.

  14. Controlling the Neutron Yield from a Small Dense Plasma Focus using Deuterium-Inert Gas Mixtures

    SciTech Connect

    Bures, B. L.; Krishnan, M.; Eshaq, Y.

    2009-01-21

    The dense plasma focus (DPF) is a well known source of neutrons when operating with deuterium. The DPF is demonstrated to scale from 10{sup 4} n/pulse at 40 kA to >10{sup 12} n/pulse at 2 MA by non-linear current scaling as described in [1], which is itself based on the simple yet elegant model developed by Lee [2]. In addition to the peak current, the gas pressure controls the neutron yield. Recent published results suggest that mixing 1-5% mass fractions of Krypton increase the neutron yield per pulse by more than 10x. In this paper we present results obtained by mixing deuterium with Helium, Neon and Argon in a 500 J dense plasma focus operating at 140 kA with a 600 ns rise time. The mass density was held constant in these experiments at the optimum (pure) deuterium mass density for producing neutrons. A typical neutron yield for a pure deuterium gas charge is 2x10{sup 6}{+-}15% n/pulse. Neutron yields in excess of 10{sup 7}{+-}10% n/pulse were observed with low mass fractions of inert gas. Time integrated optical images of the pinch, soft x-ray measurements and optical emission spectroscopy where used to examine the pinch in addition to the neutron yield monitor and the fast scintillation detector. Work supported by Domestic Nuclear Detection Office under contract HSHQDC-08-C-00020.

  15. Out-of-field doses and neutron dose equivalents for electron beams from modern Varian and Elekta linear accelerators.

    PubMed

    Cardenas, Carlos E; Nitsch, Paige L; Kudchadker, Rajat J; Howell, Rebecca M; Kry, Stephen F

    2016-01-01

    Out-of-field doses from radiotherapy can cause harmful side effects or eventually lead to secondary cancers. Scattered doses outside the applicator field, neutron source strength values, and neutron dose equivalents have not been broadly investigated for high-energy electron beams. To better understand the extent of these exposures, we measured out-of-field dose characteristics of electron applicators for high-energy electron beams on two Varian 21iXs, a Varian TrueBeam, and an Elekta Versa HD operating at various energy levels. Out-of-field dose profiles and percent depth-dose curves were measured in a Wellhofer water phantom using a Farmer ion chamber. Neutron dose was assessed using a combination of moderator buckets and gold activation foils placed on the treatment couch at various locations in the patient plane on both the Varian 21iX and Elekta Versa HD linear accelerators. Our findings showed that out-of-field electron doses were highest for the highest electron energies. These doses typically decreased with increasing distance from the field edge but showed substantial increases over some distance ranges. The Elekta linear accelerator had higher electron out-of-field doses than the Varian units examined, and the Elekta dose profiles exhibited a second dose peak about 20 to 30 cm from central-axis, which was found to be higher than typical out-of-field doses from photon beams. Electron doses decreased sharply with depth before becoming nearly constant; the dose was found to decrease to a depth of approximately E(MeV)/4 in cm. With respect to neutron dosimetry, Q values and neutron dose equivalents increased with electron beam energy. Neutron contamination from electron beams was found to be much lower than that from photon beams. Even though the neutron dose equivalent for electron beams represented a small portion of neutron doses observed under photon beams, neutron doses from electron beams may need to be considered for special cases. PMID:27455499

  16. Neutron spectra and dose equivalents calculated in tissue for high-energy radiation therapy

    SciTech Connect

    Kry, Stephen F.; Howell, Rebecca M.; Salehpour, Mohammad; Followill, David S.

    2009-04-15

    Neutrons are by-products of high-energy radiation therapy and a source of dose to normal tissues. Thus, the presence of neutrons increases a patient's risk of radiation-induced secondary cancer. Although neutrons have been thoroughly studied in air, little research has been focused on neutrons at depths in the patient where radiosensitive structures may exist, resulting in wide variations in neutron dose equivalents between studies. In this study, we characterized properties of neutrons produced during high-energy radiation therapy as a function of their depth in tissue and for different field sizes and different source-to-surface distances (SSD). We used a previously developed Monte Carlo model of an accelerator operated at 18 MV to calculate the neutron fluences, energy spectra, quality factors, and dose equivalents in air and in tissue at depths ranging from 0.1 to 25 cm. In conjunction with the sharply decreasing dose equivalent with increased depth in tissue, the authors found that the neutron energy spectrum changed drastically as a function of depth in tissue. The neutron fluence decreased gradually as the depth increased, while the average neutron energy decreased sharply with increasing depth until a depth of approximately 7.5 cm in tissue, after which it remained nearly constant. There was minimal variation in the quality factor as a function of depth. At a given depth in tissue, the neutron dose equivalent increased slightly with increasing field size and decreasing SSD; however, the percentage depth-dose equivalent curve remained constant outside the primary photon field. Because the neutron dose equivalent, fluence, and energy spectrum changed substantially with depth in tissue, we concluded that when the neutron dose equivalent is being determined at a depth within a patient, the spectrum and quality factor used should be appropriate for depth rather than for in-air conditions. Alternately, an appropriate percent depth-dose equivalent curve should be

  17. Preliminary On-Orbit Neutron Dose Equivalent and Energy Spectrum Results from the ISS-RAD Fast Neutron Detector (FND)

    NASA Technical Reports Server (NTRS)

    Semones, Edward; Leitgab, Martin

    2016-01-01

    The ISS-RAD instrument was activated on ISS on February 1st, 2016. Integrated in ISS-RAD, the Fast Neutron Detector (FND) performs, for the first time on ISS, routine and precise direct neutron measurements between 0.5 and 8 MeV. Preliminary results for neutron dose equivalent and neutron flux energy distributions from online/on-board algorithms and offline ground analyses will be shown, along with comparisons to simulated data and previously measured neutron spectral data. On-orbit data quality and pre-launch analysis validation results will be discussed as well.

  18. Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model

    PubMed Central

    Zhang, Xiaomin; Xie, Xiangdong; Qu, Decheng; Ning, Jing; Zhou, Hongmei; Pan, Jie; Yang, Guoshan

    2016-01-01

    A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10−9 MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal–ventral, ventral–dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 MeV. PMID:26661852

  19. Experimental verification of improved depth-dose distribution using hyper-thermal neutron incidence in neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2001-01-01

    We have proposed the utilization of `hyper-thermal neutrons' for neutron capture therapy (NCT) from the viewpoint of the improvement in the dose distribution in a human body. In order to verify the improved depth-dose distribution due to hyper-thermal neutron incidence, two experiments were carried out using a test-type hyper-thermal neutron generator at a thermal neutron irradiation field in Kyoto University Reactor (KUR), which is actually utilized for NCT clinical irradiation. From the free-in-air experiment for the spectrum-shift characteristics, it was confirmed that the hyper-thermal neutrons of approximately 860 K at maximum could be obtained by the generator. From the phantom experiment, the improvement effect and the controllability for the depth-dose distribution were confirmed. For example, it was found that the relative neutron depth-dose distribution was about 1 cm improved with the 860 K hyper-thermal neutron incidence, compared to the normal thermal neutron incidence.

  20. Experimental verification of improved depth-dose distribution using hyper-thermal neutron incidence in neutron capture therapy.

    PubMed

    Sakurai, Y; Kobayashi, T

    2001-01-01

    We have proposed the utilization of 'hyper-thermal neutrons' for neutron capture therapy (NCT) from the viewpoint of the improvement in the dose distribution in a human body. In order to verify the improved depth-dose distribution due to hyper-thermal neutron incidence, two experiments were carried out using a test-type hyper-thermal neutron generator at a thermal neutron irradiation field in Kyoto University Reactor (KUR), which is actually utilized for NCT clinical irradiation. From the free-in-air experiment for the spectrum-shift characteristics, it was confirmed that the hyper-thermal neutrons of approximately 860 K at maximum could be obtained by the generator. From the phantom experiment, the improvement effect and the controllability for the depth-dose distribution were confirmed. For example, it was found that the relative neutron depth-dose distribution was about 1 cm improved with the 860 K hyper-thermal neutron incidence, compared to the normal thermal neutron incidence.

  1. Calculation of the absorbed dose and dose equivalent induced by medium energy neutrons and protons and comparison with experiment

    NASA Technical Reports Server (NTRS)

    Armstrong, T. W.; Bishop, B. L.

    1972-01-01

    Monte Carlo calculations have been carried out to determine the absorbed dose and dose equivalent for 592-MeV protons incident on a cylindrical phantom and for neutrons from 580-MeV proton-Be collisions incident on a semi-infinite phantom. For both configurations, the calculated depth dependence of the absorbed dose is in good agreement with experimental data.

  2. Calibration methodology for proportional counters applied to yield measurements of a neutron burst.

    PubMed

    Tarifeño-Saldivia, Ariel; Mayer, Roberto E; Pavez, Cristian; Soto, Leopoldo

    2014-01-01

    This paper introduces a methodology for the yield measurement of a neutron burst using neutron proportional counters. This methodology is to be applied when single neutron events cannot be resolved in time by nuclear standard electronics, or when a continuous current cannot be measured at the output of the counter. The methodology is based on the calibration of the counter in pulse mode, and the use of a statistical model to estimate the number of detected events from the accumulated charge resulting from the detection of the burst of neutrons. The model is developed and presented in full detail. For the measurement of fast neutron yields generated from plasma focus experiments using a moderated proportional counter, the implementation of the methodology is herein discussed. An experimental verification of the accuracy of the methodology is presented. An improvement of more than one order of magnitude in the accuracy of the detection system is obtained by using this methodology with respect to previous calibration methods.

  3. Time-Dependent Neutron and Photon Dose-Field Analysis

    SciTech Connect

    Wooten, Hasani Omar

    2005-08-01

    A unique tool is developed that allows the user to model physical representations of complicated glovebox facilities in two dimensions and determine neutral-particle flux and ambient dose-equivalent fields throughout that geometry. The Pandemonium code, originally designed to determine flux and dose-rates only, is improved to include realistic glovebox geometries, time-dependent source and detector positions, time-dependent shielding thickness calculations, time-integrated doses, a representative criticality accident scenario based on time-dependent reactor kinetics, and more rigorous photon treatment. A primary benefit of this work has been an extensive analysis and improvement of the photon model that is not limited to the application described in this thesis. The photon model has been extended in energy range to 10 MeV to include photons from fission and new photon buildup factors have been included that account for the effects of photon buildup at slant-path thicknesses as a function of angle, where the mean free path thickness has been preserved. The overall system of codes is user-friendly and it is directly applicable to facilities such as the plutonium facility at Los Alamos National Laboratory, where high-intensity neutron and photon emitters are regularly used. The codes may be used to determine a priori doses for given work scenarios in an effort to supply dose information to process models which will in turn assist decision makers on ensuring as low as reasonably achievable (ALARA) compliance. In addition, coupling the computational results of these tools with the process model visualization tools will help to increase worker safety and radiological safety awareness.

  4. High yield neutron generator based on a high-current gasdynamic electron cyclotron resonance ion source

    SciTech Connect

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

    2015-09-07

    In present paper, an approach for high yield compact D-D neutron generator based on a high current gasdynamic electron cyclotron resonance ion source is suggested. Results on dense pulsed deuteron beam production with current up to 500 mA and current density up to 750 mA/cm{sup 2} are demonstrated. Neutron yield from D{sub 2}O and TiD{sub 2} targets was measured in case of its bombardment by pulsed 300 mA D{sup +} beam with 45 keV energy. Neutron yield density at target surface of 10{sup 9} s{sup −1} cm{sup −2} was detected with a system of two {sup 3}He proportional counters. Estimations based on obtained experimental results show that neutron yield from a high quality TiD{sub 2} target bombarded by D{sup +} beam demonstrated in present work accelerated to 100 keV could reach 6 × 10{sup 10} s{sup −1} cm{sup −2}. It is discussed that compact neutron generator with such characteristics could be perspective for a number of applications like boron neutron capture therapy, security systems based on neutron scanning, and neutronography.

  5. Time-dependent neutron and photon dose-field analysis

    NASA Astrophysics Data System (ADS)

    Wooten, Hasani Omar

    2005-11-01

    A unique tool is developed that allows the user to model physical representations of complicated glovebox facilities in two dimensions and determine neutral-particle flux and ambient dose-equivalent fields throughout that geometry. The code Pandemonium, originally designed to determine flux and dose rates only, has been improved to include realistic glovebox geometries, time-dependent source and detector positions, time-dependent shielding thickness calculations, time-integrated doses, a representative criticality accident scenario based on time-dependent reactor kinetics, and more rigorous photon treatment. The photon model has been significantly enhanced by expanding the energy range to 10 MeV to include fission photons, and by including a set of new buildup factors, the result of an extensive study into the previously unknown "purely-angular effect" on photon buildup. Purely-angular photon buildup factors are determined using discrete ordinates and coupled electron-photon cross sections to account for coherent and incoherent scattering and secondary photon effects of bremsstrahlung and florescence. Improvements to Pandemonium result in significant modeling capabilities for processing facilities using intense neutron and photon sources, and the code obtains comparable results to Monte Carlo calculations but within a fraction of the time required to run such codes as MCNPX.

  6. Two-dimensional simulations of the neutron yield in cryogenic deuterium-tritium implosions on OMEGA

    NASA Astrophysics Data System (ADS)

    Hu, S. X.; Goncharov, V. N.; Radha, P. B.; Marozas, J. A.; Skupsky, S.; Boehly, T. R.; Sangster, T. C.; Meyerhofer, D. D.; McCrory, R. L.

    2010-10-01

    Maximizing the neutron yield to obtain energy gain is the ultimate goal for inertial confinement fusion. Nonuniformities seeded by target and laser perturbations can disrupt neutron production via the Rayleigh-Taylor instability growth. To understand the effects of perturbations on the neutron yield of cryogenic DT implosions on the Omega Laser Facility [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)], two-dimensional DRACO [P. B. Radha et al., Phys. Plasmas 12, 056307 (2005)] simulations have been performed to systematically investigate each perturbation source and their combined effects on the neutron-yield performance. Two sources of nonuniformity accounted for the neutron-yield reduction in DRACO simulations: target offset from the target chamber center and laser imprinting. The integrated simulations for individual shots reproduce the experimental yield-over-clean (YOC) ratio within a factor of 2 or better. The simulated neutron-averaged ion temperatures ⟨Ti⟩ is only about 10%-15% higher than measurements. By defining the temperature-over-clean, its relationship to YOC provides an indication of how much the hot-spot volume and density are perturbed with respect to the uniform situation. Typically, the YOC in OMEGA experiments is of the order of ˜5%. The simulation results suggest that YOC can be increased to the ignition hydroequivalent level of 15%-20% (with ⟨ρR⟩=200-300 mg/cm2) by maintaining a target offset of less than 10 μm and employing beam smoothing by spectral dispersion.

  7. Neutron yield and induced radioactivity: a study of 235-MeV proton and 3-GeV electron accelerators.

    PubMed

    Hsu, Yung-Cheng; Lai, Bo-Lun; Sheu, Rong-Jiun

    2016-01-01

    This study evaluated the magnitude of potential neutron yield and induced radioactivity of two new accelerators in Taiwan: a 235-MeV proton cyclotron for radiation therapy and a 3-GeV electron synchrotron serving as the injector for the Taiwan Photon Source. From a nuclear interaction point of view, neutron production from targets bombarded with high-energy particles is intrinsically related to the resulting target activation. Two multi-particle interaction and transport codes, FLUKA and MCNPX, were used in this study. To ensure prediction quality, much effort was devoted to the associated benchmark calculations. Comparisons of the accelerators' results for three target materials (copper, stainless steel and tissue) are presented. Although the proton-induced neutron yields were higher than those induced by electrons, the maximal neutron production rates of both accelerators were comparable according to their respective beam outputs during typical operation. Activation products in the targets of the two accelerators were unexpectedly similar because the primary reaction channels for proton- and electron-induced activation are (p,pn) and (γ,n), respectively. The resulting residual activities and remnant dose rates as a function of time were examined and discussed.

  8. Neutron dose estimation via LET spectrometry using CR-39 detector for the reaction 9Be (p, n)

    PubMed Central

    Sahoo, G. S.; Tripathy, S. P.; Paul, S.; Sharma, S. D.; Sharma, S. C.; Joshi, D. S.; Bandyopadhyay, T.

    2014-01-01

    CR-39 detectors, widely used for neutron dosimetry in accelerator radiation environment, have also been applied in tissue microdosimetry by generating the linear energy transfer (LET) spectrum. In this work, the neutron dose has been estimated via LET spectrometry for 9Be (p, n) reaction which is useful for personnel monitoring around particle accelerators and accelerator based therapy facilities. Neutrons were generated by the interaction of protons of 6 different energies from 4–24 MeV with a thick Be target. The LET spectra were obtained from the major and minor radii of each track and the thickness of removed surface. From the LET spectra, the absorbed dose (DLET) and the dose equivalent (HLET) were estimated using Q-L relationship as given by International Commission on Radiological Protection (ICRP) 60. The track density in CR-39 detector and hence the neutron yield was found to be increasing with the increase in projectile (proton) energy. Similar observations were also obtained for absorbed dose (DLET) and dose equivalents (HLET). PMID:25525310

  9. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle. PMID:27058075

  10. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  11. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  12. Secondary neutron dose measurement for proton eye treatment using an eye snout with a borated neutron absorber

    PubMed Central

    2013-01-01

    Background We measured and assessed ways to reduce the secondary neutron dose from a system for proton eye treatment. Methods Proton beams of 60.30 MeV were delivered through an eye-treatment snout in passive scattering mode. Allyl diglycol carbonate (CR-39) etch detectors were used to measure the neutron dose in the external field at 0.00, 1.64, and 6.00 cm depths in a water phantom. Secondary neutron doses were measured and compared between those with and without a high-hydrogen–boron-containing block. In addition, the neutron energy and vertices distribution were obtained by using a Geant4 Monte Carlo simulation. Results The ratio of the maximum neutron dose equivalent to the proton absorbed dose (H(10)/D) at 2.00 cm from the beam field edge was 8.79 ± 1.28 mSv/Gy. The ratio of the neutron dose equivalent to the proton absorbed dose with and without a high hydrogen-boron containing block was 0.63 ± 0.06 to 1.15 ± 0.13 mSv/Gy at 2.00 cm from the edge of the field at depths of 0.00, 1.64, and 6.00 cm. Conclusions We found that the out-of-field secondary neutron dose in proton eye treatment with an eye snout is relatively small, and it can be further reduced by installing a borated neutron absorbing material. PMID:23866307

  13. In-phantom neutron dose distribution for bladder cancer cases treated with high-energy photons

    NASA Astrophysics Data System (ADS)

    Khaled, N. E.; Attalla, E. M.; Ammar, H.; Khalil, W.

    2011-06-01

    This work presents an estimation of the neutron dose distribution for common bladder cancer cases treated with high-energy photons of 15 MV therapy accelerators. Neutron doses were measured in an Alderson phantom, using TLD 700 and 600 thermoluminescence dosimeters, resembling bladder cancer cases treated with high-energy photons from 15 MV LINAC and having a treatment plan using the four-field pelvic box technique. Thermal neutron dose distribution in the target area and the surrounding tissue was estimated. The sensitivity of all detectors for both gamma and neutrons was estimated and used for correction of the TL reading. TLD detectors were irradiated with a Co60 gamma standard source and thermal neutrons at the irradiation facility of the National Institute for Standards (NIS). The TL to dose conversion factor was estimated in terms of both Co60 neutron equivalent dose and thermal neutron dose. The dose distribution of photo-neutrons throughout each target was estimated and presented in three-dimensional charts and isodose curves. The distribution was found to be non-isotropic through the target. It varied from a minimum of 0.23 mSv/h to a maximum of 2.07 mSv/h at 6 cm off-axis. The mean neutron dose equivalent was found to be 0.63 mSv/h, which agrees with other published literature. The estimated average neutron equivalent to the bladder per administered therapeutic dose was found to be 0.39 mSv Gy-1, which is also in good agreement with published literature. As a consequence of a complete therapeutic treatment of 50 Gy high-energy photons at 15 MV, the total thermal neutron equivalent dose to the abdomen was found to be about 0.012 Sv.

  14. High-dose neutron irradiation performance of dielectric mirrors

    DOE PAGES

    Nimishakavi Anantha Phani Kiran Kumar; Leonard, Keith J.; Jellison, Jr., Gerald Earle; Snead, Lance Lewis

    2015-05-01

    The study presents the high-dose behavior of dielectric mirrors specifically engineered for radiation-tolerance: alternating layers of Al2O3/SiO2 and HfO2/SiO2 were grown on sapphire substrates and exposed to neutron doses of 1 and 4 dpa at 458 10K in the High Flux Isotope Reactor (HFIR). In comparison to previously reported results, these higher doses of 1 and 4 dpa results in a drastic drop in optical reflectance, caused by a failure of the multilayer coating. HfO2/SiO2 mirrors failed completely when exposed to 1 dpa, whereas the reflectance of Al2O3/SiO2 mirrors reduced to 44%, eventually failing at 4 dpa. Transmission electron microscopymore » (TEM) observation of the Al2O3/SiO2 specimens showed SiO2 layer defects which increases size with irradiation dose. The typical size of each defect was 8 nm in 1 dpa and 42 nm in 4 dpa specimens. Buckling type delamination of the interface between the substrate and first layer was typically observed in both 1 and 4 dpa HfO2/SiO2 specimens. Composition changes across the layers were measured in high resolution scanning-TEM mode using energy dispersive spectroscopy. A significant interdiffusion between the film layers was observed in Al2O3/SiO2 mirror, though less evident in HfO2/SiO2 system. Lastly, the ultimate goal of this work is the provide insight into the radiation-induced failure mechanisms of these mirrors.« less

  15. High-dose neutron irradiation performance of dielectric mirrors

    SciTech Connect

    Nimishakavi Anantha Phani Kiran Kumar; Leonard, Keith J.; Jellison, Jr., Gerald Earle; Snead, Lance Lewis

    2015-05-01

    The study presents the high-dose behavior of dielectric mirrors specifically engineered for radiation-tolerance: alternating layers of Al2O3/SiO2 and HfO2/SiO2 were grown on sapphire substrates and exposed to neutron doses of 1 and 4 dpa at 458 10K in the High Flux Isotope Reactor (HFIR). In comparison to previously reported results, these higher doses of 1 and 4 dpa results in a drastic drop in optical reflectance, caused by a failure of the multilayer coating. HfO2/SiO2 mirrors failed completely when exposed to 1 dpa, whereas the reflectance of Al2O3/SiO2 mirrors reduced to 44%, eventually failing at 4 dpa. Transmission electron microscopy (TEM) observation of the Al2O3/SiO2 specimens showed SiO2 layer defects which increases size with irradiation dose. The typical size of each defect was 8 nm in 1 dpa and 42 nm in 4 dpa specimens. Buckling type delamination of the interface between the substrate and first layer was typically observed in both 1 and 4 dpa HfO2/SiO2 specimens. Composition changes across the layers were measured in high resolution scanning-TEM mode using energy dispersive spectroscopy. A significant interdiffusion between the film layers was observed in Al2O3/SiO2 mirror, though less evident in HfO2/SiO2 system. Lastly, the ultimate goal of this work is the provide insight into the radiation-induced failure mechanisms of these mirrors.

  16. Neutron and gamma dose and spectra measurements on the Little Boy replica

    SciTech Connect

    Hoots, S.; Wadsworth, D.

    1984-06-01

    The radiation-measurement team of the Weapons Engineering Division at Lawrence Livermore National Laboratory (LLNL) measured neutron and gamma dose and spectra on the Little Boy replica at Los Alamos National Laboratory (LANL) in April 1983. This assembly is a replica of the gun-type atomic bomb exploded over Hiroshima in 1945. These measurements support the National Academy of Sciences Program to reassess the radiation doses due to atomic bomb explosions in Japan. Specifically, the following types of information were important: neutron spectra as a function of geometry, gamma to neutron dose ratios out to 1.5 km, and neutron attenuation in the atmosphere. We measured neutron and gamma dose/fission from close-in to a kilometer out, and neutron and gamma spectra at 90 and 30/sup 0/ close-in. This paper describes these measurements and the results. 12 references, 13 figures, 5 tables.

  17. Measurement of neutron yield by 62 MeV proton beam on a thick beryllium target

    NASA Astrophysics Data System (ADS)

    Osipenko, M.; Ripani, M.; Alba, R.; Ricco, G.; Schillaci, M.; Barbagallo, M.; Boccaccio, P.; Celentano, A.; Colonna, N.; Cosentino, L.; Del Zoppo, A.; Di Pietro, A.; Esposito, J.; Figuera, P.; Finocchiaro, P.; Kostyukov, A.; Maiolino, C.; Santonocito, D.; Scuderi, V.; Viberti, C. M.

    2013-09-01

    The design of a low-power prototype of neutron amplifier recently proposed within the INFN-E project indicated the need for more accurate data on the neutron yield produced by a proton beam with energy of about 70 MeV impinging on a thick beryllium target. Such measurement was performed at the LNS superconducting cyclotron, covering a wide angular range from 0° to 150° and a complete neutron energy interval from thermal to beam energy. Neutrons with energy above 0.5 MeV were measured by liquid scintillators exploiting their time of flight to determine the kinetic energy. For lower energy neutrons, down to thermal energy, a 3He detector was used. The obtained data are in good agreement with previous measurements at 0° using 66 MeV proton beam, covering neutron energies >10 MeV, as well as with measurements at few selected angles using protons of 46, 55 and 113 MeV energy. The present results extend the neutron yield data in the 60-70 MeV beam energy range. A comparison of measured yields to MCNP, FLUKA and Geant4 Monte Carlo simulations was performed.

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

    PubMed

    Delfin, A; Paredes, L C; Zambrano, F; Guzmán-Rincón, J; Ureña-Nuñez, F

    2001-12-01

    A method to obtain the absorbed dose in Drosophila melanogaster irradiated in the thermal column facility of the Triga Mark III Reactor has been developed. The method is based on the measurements of neutron activation of gold foils produced by neutron capture to obtain the neutron fluxes. These fluxes, combined with the calculations of kinetic energy released per unit mass, enables one to obtain the absorbed doses in Drosophila melanogaster. PMID:11761104

  19. Impact of the Revised 10 CFR 835 on the Neutron Dose Rates at LLNL

    SciTech Connect

    Radev, R

    2009-01-13

    In June 2007, 10 CFR 835 [1] was revised to include new radiation weighting factors for neutrons, updated dosimetric models, and dose terms consistent with the newer ICRP recommendations. A significant aspect of the revised 10 CFR 835 is the adoption of the recommendations outlined in ICRP-60 [2]. The recommended new quantities demand a review of much of the basic data used in protection against exposure to sources of ionizing radiation. The International Commission on Radiation Units and Measurements has defined a number of quantities for use in personnel and area monitoring [3,4,5] including the ambient dose equivalent H*(d) to be used for area monitoring and instrument calibrations. These quantities are used in ICRP-60 and ICRP-74. This report deals only with the changes in the ambient dose equivalent and ambient dose rate equivalent for neutrons as a result of the implementation of the revised 10 CFR 835. In the report, the terms neutron dose and neutron dose rate will be used for convenience for ambient neutron dose and ambient neutron dose rate unless otherwise stated. This report provides a qualitative and quantitative estimate of how much the neutron dose rates at LLNL will change with the implementation of the revised 10 CFR 835. Neutron spectra and dose rates from selected locations at the LLNL were measured with a high resolution spectroscopic neutron dose rate system (ROSPEC) as well as with a standard neutron rem meter (a.k.a., a remball). The spectra obtained at these locations compare well with the spectra from the Radiation Calibration Laboratory's (RCL) bare californium source that is currently used to calibrate neutron dose rate instruments. The measurements obtained from the high resolution neutron spectrometer and dose meter ROSPEC and the NRD dose meter compare within the range of {+-}25%. When the new radiation weighting factors are adopted with the implementation of the revised 10 CFR 835, the measured dose rates will increase by up to 22

  20. Implementation of an Analytical Model for Leakage Neutron Equivalent Dose in a Proton Radiotherapy Planning System

    PubMed Central

    Eley, John; Newhauser, Wayne; Homann, Kenneth; Howell, Rebecca; Schneider, Christopher; Durante, Marco; Bert, Christoph

    2015-01-01

    Equivalent dose from neutrons produced during proton radiotherapy increases the predicted risk of radiogenic late effects. However, out-of-field neutron dose is not taken into account by commercial proton radiotherapy treatment planning systems. The purpose of this study was to demonstrate the feasibility of implementing an analytical model to calculate leakage neutron equivalent dose in a treatment planning system. Passive scattering proton treatment plans were created for a water phantom and for a patient. For both the phantom and patient, the neutron equivalent doses were small but non-negligible and extended far beyond the therapeutic field. The time required for neutron equivalent dose calculation was 1.6 times longer than that required for proton dose calculation, with a total calculation time of less than 1 h on one processor for both treatment plans. Our results demonstrate that it is feasible to predict neutron equivalent dose distributions using an analytical dose algorithm for individual patients with irregular surfaces and internal tissue heterogeneities. Eventually, personalized estimates of neutron equivalent dose to organs far from the treatment field may guide clinicians to create treatment plans that reduce the risk of late effects. PMID:25768061

  1. Development of a dual phantom technique for measuring the fast neutron component of dose in boron neutron capture therapy

    SciTech Connect

    Sakurai, Yoshinori Tanaka, Hiroki; Kondo, Natsuko; Kinashi, Yuko; Suzuki, Minoru; Masunaga, Shinichiro; Ono, Koji; Maruhashi, Akira

    2015-11-15

    Purpose: Research and development of various accelerator-based irradiation systems for boron neutron capture therapy (BNCT) is underway throughout the world. Many of these systems are nearing or have started clinical trials. Before the start of treatment with BNCT, the relative biological effectiveness (RBE) for the fast neutrons (over 10 keV) incident to the irradiation field must be estimated. Measurements of RBE are typically performed by biological experiments with a phantom. Although the dose deposition due to secondary gamma rays is dominant, the relative contributions of thermal neutrons (below 0.5 eV) and fast neutrons are virtually equivalent under typical irradiation conditions in a water and/or acrylic phantom. Uniform contributions to the dose deposited from thermal and fast neutrons are based in part on relatively inaccurate dose information for fast neutrons. This study sought to improve the accuracy in the dose estimation for fast neutrons by using two phantoms made of different materials in which the dose components can be separated according to differences in the interaction cross sections. The development of a “dual phantom technique” for measuring the fast neutron component of dose is reported. Methods: One phantom was filled with pure water. The other phantom was filled with a water solution of lithium hydroxide (LiOH) capitalizing on the absorbing characteristics of lithium-6 (Li-6) for thermal neutrons. Monte Carlo simulations were used to determine the ideal mixing ratio of Li-6 in LiOH solution. Changes in the depth dose distributions for each respective dose component along the central beam axis were used to assess the LiOH concentration at the 0, 0.001, 0.01, 0.1, 1, and 10 wt. % levels. Simulations were also performed with the phantom filled with 10 wt. % {sup 6}LiOH solution for 95%-enriched Li-6. A phantom was constructed containing 10 wt. % {sup 6}LiOH solution based on the simulation results. Experimental characterization of the

  2. Dose evaluation of boron neutron capture synovectomy using the THOR epithermal neutron beam: a feasibility study

    NASA Astrophysics Data System (ADS)

    Wu, Jay; Chang, Shu-Jun; Chuang, Keh-Shih; Hsueh, Yen-Wan; Yeh, Kuan-Chuan; Wang, Jeng-Ning; Tsai, Wen-Pin

    2007-03-01

    Rheumatoid arthritis is one of the most common epidemic diseases in the world. For some patients, the treatment with steroids or nonsteroidal anti-inflammatory drugs is not effective, thus necessitating physical removal of the inflamed synovium. Alternative approaches other than surgery will provide appropriate disease control and improve the patient's quality of life. In this research, we evaluated the feasibility of conducting boron neutron capture synovectomy (BNCS) with the Tsing Hua open-pool reactor (THOR) as a neutron source. Monte Carlo simulations were performed with arthritic joint models and uncertainties were within 5%. The collimator, reflector and boron concentration were optimized to reduce the treatment time and normal tissue doses. For the knee joint, polyethylene with 40%-enriched Li2CO3 was used as the collimator material, and a rear reflector of 15 cm thick graphite and side reflector of 10 cm thick graphite were chosen. The optimized treatment time was 5.4 min for the parallel-opposed irradiation. For the finger joint, polymethyl methacrylate was used as the reflector material. The treatment time can be reduced to 3.1 min, while skin and bone doses can be effectively reduced by approximately 9% compared with treatment using the graphite reflector. We conclude that using THOR as a treatment modality for BNCS could be a feasible alternative in clinical practice.

  3. Fusion-neutron-yield, activation measurements at the Z accelerator: Design, analysis, and sensitivity

    SciTech Connect

    Hahn, K. D. Ruiz, C. L.; Fehl, D. L.; Chandler, G. A.; Knapp, P. F.; Smelser, R. M.; Torres, J. A.; Cooper, G. W.; Nelson, A. J.; Leeper, R. J.

    2014-04-15

    We present a general methodology to determine the diagnostic sensitivity that is directly applicable to neutron-activation diagnostics fielded on a wide variety of neutron-producing experiments, which include inertial-confinement fusion (ICF), dense plasma focus, and ion beam-driven concepts. This approach includes a combination of several effects: (1) non-isotropic neutron emission; (2) the 1/r{sup 2} decrease in neutron fluence in the activation material; (3) the spatially distributed neutron scattering, attenuation, and energy losses due to the fielding environment and activation material itself; and (4) temporally varying neutron emission. As an example, we describe the copper-activation diagnostic used to measure secondary deuterium-tritium fusion-neutron yields on ICF experiments conducted on the pulsed-power Z Accelerator at Sandia National Laboratories. Using this methodology along with results from absolute calibrations and Monte Carlo simulations, we find that for the diagnostic configuration on Z, the diagnostic sensitivity is 0.037% ± 17% counts/neutron per cm{sup 2} and is ∼ 40% less sensitive than it would be in an ideal geometry due to neutron attenuation, scattering, and energy-loss effects.

  4. SU-E-T-566: Neutron Dose Cloud Map for Compact ProteusONE Proton Therapy

    SciTech Connect

    Syh, J; Patel, B; Syh, J; Rosen, L; Wu, H

    2015-06-15

    Purpose: To establish the base line of neutron cloud during patient treatment in our new compact Proteus One proton pencil beam scanning (PBS) system with various beam delivery gantry angles, with or without range shifter (RS) at different body sites. Pencil beam scanning is an emerging treatment technique, for the concerns of neutron exposure, this study is to evaluate the neutron dose equivalent per given delivered dose under various treatment conditions at our proton therapy center. Methods: A wide energy neutron dose equivalent detector (SWENDI-II, Thermo Scientific, MA) was used for neutron dose measurements. It was conducted in the proton therapy vault during beam was on. The measurement location was specifically marked in order to obtain the equivalent dose of neutron activities (H). The distances of 100, 150 and 200 cm at various locations are from the patient isocenter. The neutron dose was measured of proton energy layers, # of spots, maximal energy range, modulation width, field radius, gantry angle, snout position and delivered dose in CGE. The neutron dose cloud is reproducible and is useful for the future reference. Results: When distance increased the neutron equivalent dose (H) reading did not decrease rapidly with changes of proton energy range, modulation width or spot layers. For cranial cases, the average mSv/CGE was about 0.02 versus 0.032 for pelvis cases. RS will induce higher H to be 0.10 mSv/CGE in average. Conclusion: From this study, neutron per dose ratio (mSv/CGE) slightly depends upon various treatment parameters for pencil beams. For similar treatment conditions, our measurement demonstrates this value for pencil beam scanning beam has lowest than uniform scanning or passive scattering beam with a factor of 5. This factor will be monitored continuously for other upcoming treatment parameters in our facility.

  5. Calibration methodology for proportional counters applied to yield measurements of a neutron burst

    NASA Astrophysics Data System (ADS)

    Tarifeño-Saldivia, Ariel; Mayer, Roberto E.; Pavez, Cristian; Soto, Leopoldo

    2015-03-01

    This work introduces a methodology for the yield measurement of a neutron burst using neutron proportional counters. The methodology is based on the calibration of the counter in pulse mode, and the use of a statistical model to estimate the number of detected events from the accumulated charge resulting from detection of the burst of neutrons. An improvement of more than one order of magnitude in the accuracy of a paraffin wax moderated 3He-filled tube is obtained by using this methodology with respect to previous calibration methods.

  6. The neutron dose conversion coefficients calculation in human tooth enamel in an anthropomorphic phantom.

    PubMed

    Khailov, A M; Ivannikov, A I; Skvortsov, V G; Stepanenko, V F; Tsyb, A F; Trompier, F; Hoshi, M

    2010-02-01

    In the present study, MCNP4B simulation code is used to simulate neutron and photon transport. It gives the conversion coefficients that relate neutron fluence to the dose in tooth enamel (molars and pre-molars only) for 20 energy groups of monoenergetic neutrons with energies from 10-9 to 20 MeV for five different irradiation geometries. The data presented are intended to provide the basis for connection between EPR dose values and standard protection quantities defined in ICRP Publication 74. The results of the calculations for critical organs were found to be consistent with ICRP data, with discrepancies generally less than 10% for the fast neutrons. The absorbed dose in enamel was found to depend strongly on the incident neutron energy for neutrons over 10 keV. The dependence of the data on the irradiation geometry is also shown. Lower bound estimates of enamel radiation sensitivity to neutrons were made using obtained coefficients for the secondary photons. Depending on neutron energy, tooth enamel was shown to register 10-120% of the total neutron dose in the human body in the case of pure neutron exposure and AP irradiation geometry. PMID:20065707

  7. Peripheral photon and neutron doses from prostate cancer external beam irradiation.

    PubMed

    Bezak, Eva; Takam, Rundgham; Marcu, Loredana G

    2015-12-01

    Peripheral photon and neutron doses from external beam radiotherapy (EBRT) are associated with increased risk of carcinogenesis in the out-of-field organs; thus, dose estimations of secondary radiation are imperative. Peripheral photon and neutron doses from EBRT of prostate carcinoma were measured in Rando phantom. (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P glass-rod thermoluminescence dosemeters (TLDs) were inserted in slices of a Rando phantom followed by exposure to 80 Gy with 18-MV photon four-field 3D-CRT technique. The TLDs were calibrated using 6- and 18-MV X-ray beam. Neutron dose equivalents measured with CR-39 etch-track detectors were used to derive readout-to-neutron dose conversion factor for (6)LiF:Mg,Cu,P TLDs. Average neutron dose equivalents per 1 Gy of isocentre dose were 3.8±0.9 mSv Gy(-1) for thyroid and 7.0±5.4 mSv Gy(-1) for colon. For photons, the average dose equivalents per 1 Gy of isocentre dose were 0.2±0.1 mSv Gy(-1) for thyroid and 8.1±9.7 mSv Gy(-1) for colon. Paired (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P TLDs can be used to measure photon and neutron doses simultaneously. Organs in close proximity to target received larger doses from photons than those from neutrons whereas distally located organs received higher neutron versus photon dose.

  8. Improvement of dose distribution by central beam shielding in boron neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Sakurai, Yoshinori; Ono, Koji

    2007-12-01

    Since boron neutron capture therapy (BNCT) with epithermal neutron beams started at the Kyoto University Reactor (KUR) in June 2002, nearly 200 BNCT treatments have been carried out. The epithermal neutron irradiation significantly improves the dose distribution, compared with the previous irradiation mainly using thermal neutrons. However, the treatable depth limit still remains. One effective technique to improve the limit is the central shield method. Simulations were performed for the incident neutron energies and the annular components of the neutron source. It was clear that thermal neutron flux distribution could be improved by decreasing the lower energy neutron component and the inner annular component of the incident beam. It was found that a central shield of 4-6 cm diameter and 10 mm thickness is effective for the 12 cm diameter irradiation field. In BNCT at KUR, the depth dose distribution can be much improved by the central shield method, resulting in a relative increase of the dose at 8 cm depth by about 30%. In addition to the depth dose distribution, the depth dose profile is also improved. As the dose rate in the central area is reduced by the additional shielding, the necessary irradiation time, however, increases by about 30% compared to normal treatment.

  9. Evaluation of H*(10) using the developed spherical type neutron dose monitor.

    PubMed

    Bhuiya, S H; Yamanishi, H; Uda, T

    2010-10-01

    An instrument for evaluating the neutron ambient dose equivalent has been developed. It has the characteristic of uniform response to wide energy of neutrons. The monitor is four-layered spherically shaped, based on moderation and absorption of neutrons. Neutron dose can be evaluated from the linear combination of three specific responses of thermoluminescent dosimeters (TLDs), which are located at three depths in the moderator. TLDs were arranged between layers of two consecutive depths on 12 radial axes at even intervals so that the monitor is equally sensitive to all directions of neutrons. In order to verify the usefulness of dose evaluation by the monitor, irradiation experiments were conducted at the FRS, JAEA. The D2O-moderated 252Cf was used for the calibration of the monitor. Experiments were also conducted by using two neutron sources of 252Cf bare and 241Am-Be. As a result, the evaluated dose for each irradiation was obtained close to the actual irradiated dose. It was confirmed that the method of dose evaluation by the developed monitor can be applied to practical neutron fields where the distance of neutron source is unknown.

  10. Absolute calibration method for laser megajoule neutron yield measurement by activation diagnostics

    NASA Astrophysics Data System (ADS)

    Landoas, Olivier; Yu Glebov, Vladimir; Rossé, Bertrand; Briat, Michelle; Disdier, Laurent; Sangster, Thomas C.; Duffy, Tim; Marmouget, Jean Gabriel; Varignon, Cyril; Ledoux, Xavier; Caillaud, Tony; Thfoin, Isabelle; Bourgade, Jean-Luc

    2011-07-01

    The laser megajoule (LMJ) and the National Ignition Facility (NIF) plan to demonstrate thermonuclear ignition using inertial confinement fusion (ICF). The neutron yield is one of the most important parameters to characterize ICF experiment performance. For decades, the activation diagnostic was chosen as a reference at ICF facilities and is now planned to be the first nuclear diagnostic on LMJ, measuring both 2.45 MeV and 14.1 MeV neutron yields. Challenges for the activation diagnostic development are absolute calibration, accuracy, range requirement, and harsh environment. At this time, copper and zirconium material are identified for 14.1 MeV neutron yield measurement and indium material for 2.45 MeV neutrons. A series of calibrations were performed at Commissariat à l'Energie Atomique (CEA) on a Van de Graff facility to determine activation diagnostics efficiencies and to compare them with results from calculations. The CEA copper activation diagnostic was tested on the OMEGA facility during DT implosion. Experiments showed that CEA and Laboratory for Laser Energetics (LLE) diagnostics agree to better than 1% on the neutron yield measurement, with an independent calibration for each system. Also, experimental sensitivities are in good agreement with simulations and allow us to scale activation diagnostics for the LMJ measurement range.

  11. Study of asymmetric fission yield behavior from neutron-deficient Hg isotope

    SciTech Connect

    Perkasa, Y. S.; Waris, A. Kurniadi, R. Su'ud, Z.

    2014-09-30

    A study of asymmetric fission yield behavior from a neutron-deficient Hg isotope has been conducted. The fission yield calculation of the neutron-deficient Hg isotope using Brownian Metropolis shape had showed unusual result at decreasing energy. In this paper, this interesting feature will be validated by using nine degree of scission shapes parameterization from Brosa model that had been implemented in TALYS nuclear reaction code. This validation is intended to show agreement between both model and the experiment result. The expected result from these models considered to be different due to dynamical properties that implemented in both models.

  12. Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

    SciTech Connect

    Verbeke, Jerome M.

    1999-12-14

    A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only.

  13. Secondary Neutron Doses to Pediatric Patients During Intracranial Proton Therapy: Monte Carlo Simulation of the Neutron Energy Spectrum and its Organ Doses.

    PubMed

    Matsumoto, Shinnosuke; Koba, Yusuke; Kohno, Ryosuke; Lee, Choonsik; Bolch, Wesley E; Kai, Michiaki

    2016-04-01

    Proton therapy has the physical advantage of a Bragg peak that can provide a better dose distribution than conventional x-ray therapy. However, radiation exposure of normal tissues cannot be ignored because it is likely to increase the risk of secondary cancer. Evaluating secondary neutrons generated by the interaction of the proton beam with the treatment beam-line structure is necessary; thus, performing the optimization of radiation protection in proton therapy is required. In this research, the organ dose and energy spectrum were calculated from secondary neutrons using Monte Carlo simulations. The Monte Carlo code known as the Particle and Heavy Ion Transport code System (PHITS) was used to simulate the transport proton and its interaction with the treatment beam-line structure that modeled the double scattering body of the treatment nozzle at the National Cancer Center Hospital East. The doses of the organs in a hybrid computational phantom simulating a 5-y-old boy were calculated. In general, secondary neutron doses were found to decrease with increasing distance to the treatment field. Secondary neutron energy spectra were characterized by incident neutrons with three energy peaks: 1×10, 1, and 100 MeV. A block collimator and a patient collimator contributed significantly to organ doses. In particular, the secondary neutrons from the patient collimator were 30 times higher than those from the first scatter. These results suggested that proactive protection will be required in the design of the treatment beam-line structures and that organ doses from secondary neutrons may be able to be reduced. PMID:26910030

  14. Secondary Neutron Doses to Pediatric Patients During Intracranial Proton Therapy: Monte Carlo Simulation of the Neutron Energy Spectrum and its Organ Doses.

    PubMed

    Matsumoto, Shinnosuke; Koba, Yusuke; Kohno, Ryosuke; Lee, Choonsik; Bolch, Wesley E; Kai, Michiaki

    2016-04-01

    Proton therapy has the physical advantage of a Bragg peak that can provide a better dose distribution than conventional x-ray therapy. However, radiation exposure of normal tissues cannot be ignored because it is likely to increase the risk of secondary cancer. Evaluating secondary neutrons generated by the interaction of the proton beam with the treatment beam-line structure is necessary; thus, performing the optimization of radiation protection in proton therapy is required. In this research, the organ dose and energy spectrum were calculated from secondary neutrons using Monte Carlo simulations. The Monte Carlo code known as the Particle and Heavy Ion Transport code System (PHITS) was used to simulate the transport proton and its interaction with the treatment beam-line structure that modeled the double scattering body of the treatment nozzle at the National Cancer Center Hospital East. The doses of the organs in a hybrid computational phantom simulating a 5-y-old boy were calculated. In general, secondary neutron doses were found to decrease with increasing distance to the treatment field. Secondary neutron energy spectra were characterized by incident neutrons with three energy peaks: 1×10, 1, and 100 MeV. A block collimator and a patient collimator contributed significantly to organ doses. In particular, the secondary neutrons from the patient collimator were 30 times higher than those from the first scatter. These results suggested that proactive protection will be required in the design of the treatment beam-line structures and that organ doses from secondary neutrons may be able to be reduced.

  15. Clinical applications of IMRT to adenocarcinoma of the prostate: Portal dose verification and intensity modulated neutron radiotherapy

    SciTech Connect

    Santanam, Lakshmi

    2005-01-01

    Intensity modulated radiotherapy (IMRT) provides an improvement in the conformality of radiotherapy dose distributions. Its application to photon radiotherapy for prostate adenocarcinoma is well established. A quality assurance tool for verifying photon IMRT treatment and the potential application of intensity modulation to neutron radiotherapy (IMNRT) to prostate cancer are investigated here. This study evaluates the use of an amorphous silicon flat panel imager for dose verification of photon IMRT fields. Various correction factors were developed to allow accurate estimation of the absorbed dose using this portal imager. The ratio of the dose measured with the portal imager to that measured using an ionization chamber was found to be 0.991{+-}0.026 for 23 measured IMRT fields. The study also yielded an accurate estimate of the relative beamlet intensity (fluence) at the plane of the detector. The raw difference between the relative beamlet intensity predicted by the EPID and that of the planning system for 23 IMRT fields was found to be -0.65{+-}2.69. These results demonstrate the capabilities of this imager as a robust IMRT quality assurance tool. An in-house optimization algorithm was used to optimize forward planned segments for the treatment of prostate cancer using IMNRT. The applicability of two different algorithms was investigated for IMNRT dose calculation, namely, the differential scatter air ratio (DSAR) and the finite size pencil beam (FSPB) algorithms. Measured profiles and absolute point doses were compared to results calculated by the treatment planning system. Dual ion-chamber measurements were performed to determine the individual neutron and gamma doses and to estimate the whole body dose equivalent. IMNRT plans retrospectively calculated for five prostate cancer patients provided dose distributions superior to conventional fast neutron therapy. When normalized to provide equivalent target coverage, the volume of the rectum and bladder receiving

  16. Effect of driver impedance on dense plasma focus Z-pinch neutron yield

    SciTech Connect

    Sears, Jason E-mail: schmidt36@llnl.gov; Link, Anthony E-mail: schmidt36@llnl.gov; Schmidt, Andrea E-mail: schmidt36@llnl.gov; Welch, Dale

    2014-12-15

    The Z-pinch phase of a dense plasma focus (DPF) heats the plasma by rapid compression and accelerates ions across its intense electric fields, producing neutrons through both thermonuclear and beam-target fusion. Driver characteristics have empirically been shown to affect performance, as measured by neutron yield per unit of stored energy. We are exploring the effect of driver characteristics on DPF performance using particle-in-cell (PIC) simulations of a kJ scale DPF. In this work, our PIC simulations are fluid for the run-down phase and transition to fully kinetic for the pinch phase, capturing kinetic instabilities, anomalous resistivity, and beam formation during the pinch. The anode-cathode boundary is driven by a circuit model of the capacitive driver, including system inductance, the load of the railgap switches, the guard resistors, and the coaxial transmission line parameters. It is known that the driver impedance plays an important role in the neutron yield: first, it sets the peak current achieved at pinch time; and second, it affects how much current continues to flow through the pinch when the pinch inductance and resistance suddenly increase. Here we show from fully kinetic simulations how total neutron yield depends on the impedance of the driver and the distributed parameters of the transmission circuit. Direct comparisons between the experiment and simulations enhance our understanding of these plasmas and provide predictive design capability for neutron source applications.

  17. Effect of driver impedance on dense plasma focus Z-pinch neutron yield

    NASA Astrophysics Data System (ADS)

    Sears, Jason; Link, Anthony; Schmidt, Andrea; Welch, Dale

    2014-12-01

    The Z-pinch phase of a dense plasma focus (DPF) heats the plasma by rapid compression and accelerates ions across its intense electric fields, producing neutrons through both thermonuclear and beam-target fusion. Driver characteristics have empirically been shown to affect performance, as measured by neutron yield per unit of stored energy. We are exploring the effect of driver characteristics on DPF performance using particle-in-cell (PIC) simulations of a kJ scale DPF. In this work, our PIC simulations are fluid for the run-down phase and transition to fully kinetic for the pinch phase, capturing kinetic instabilities, anomalous resistivity, and beam formation during the pinch. The anode-cathode boundary is driven by a circuit model of the capacitive driver, including system inductance, the load of the railgap switches, the guard resistors, and the coaxial transmission line parameters. It is known that the driver impedance plays an important role in the neutron yield: first, it sets the peak current achieved at pinch time; and second, it affects how much current continues to flow through the pinch when the pinch inductance and resistance suddenly increase. Here we show from fully kinetic simulations how total neutron yield depends on the impedance of the driver and the distributed parameters of the transmission circuit. Direct comparisons between the experiment and simulations enhance our understanding of these plasmas and provide predictive design capability for neutron source applications.

  18. Measuring neutron yield and ρR anisotropies with activation foils at the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Bleuel, D. L.; Bernstein, L. A.; Bionta, R. M.; Cooper, G. W.; Drury, O. B.; Hagmann, C. A.; Knittel, K. M.; Leeper, R. J.; Ruiz, C. L.; Schneider, D. H. G.; Yeamans, C. B.

    2013-11-01

    Neutron yields at the National Ignition Facility (NIF) are measured with a suite of diagnostics, including activation of ˜20-200 g samples of materials undergoing a variety of energy-dependent neutron reactions. Indium samples were mounted on the end of a Diagnostic Instrument Manipulator (DIM), 25-50 cm from the implosion, to measure 2.45 MeV D-D fusion neutron yield. The 336.2 keV gamma rays from the 4.5 hour isomer of 115mIn produced by (n,n') reactions are counted in high-purity germanium detectors. For capsules producing D-T fusion reactions, zirconium and copper are activated via (n,2n) reactions at various locations around the target chamber and bay, measuring the 14 MeV neutron yield to accuracies on order of 7%. By mounting zirconium samples on ports at nine locations around the NIF chamber, anisotropies in the primary neutron emission due to fuel areal density asymmetries can be measured to a relative precision of 3%.

  19. Mutations induced in Tradescantia by small doses of X-rays and neutrons - Analysis of dose-response curves.

    NASA Technical Reports Server (NTRS)

    Sparrow, A. H.; Underbrink, A. G.; Rossi, H. H.

    1972-01-01

    Dose-response curves for pink somatic mutations in Tradescantia stamen hairs were analyzed after neutron and X-ray irradiation with doses ranging from a fraction of a rad to the region of saturation. The dose-effect relation for neutrons indicates a linear dependence from 0.01 to 8 rads; between 0.25 and 5 rads, a linear dependence is indicated for X-rays also. As a consequence the relative biological effectiveness reaches a constant value (about 50) at low doses. The observations are in good agreement with the predictions of the theory of dual radiation action and support its interpretation of the effects of radiation on higher organisms. The doubling dose of X-rays was found to be nearly 1 rad.

  20. Cosmic radiation dose in aircraft--a neutron track etch detector.

    PubMed

    Vuković, B; Radolić, V; Miklavcić, I; Poje, M; Varga, M; Planinić, J

    2007-01-01

    Cosmic radiation bombards us at high altitude by ionizing particles. The radiation environment is a complex mixture of charged particles of solar and galactic origin, as well as of secondary particles produced in interaction of the galactic cosmic particles with the nuclei of atmosphere of the Earth. The radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. The non-neutron component of cosmic radiation dose aboard ATR 42 and A 320 aircrafts (flight level of 8 and 11 km, respectively) was measured with TLD-100 (LiF:Mg,Ti) detectors and the Mini 6100 semiconductor dosimeter. The estimated occupational effective dose for the aircraft crew (A 320) working 500 h per year was 1.64 mSv. Other experiments, or dose rate measurements with the neutron dosimeter, consisting of LR-115 track detector and boron foil BN-1 or 10B converter, were performed on five intercontinental flights. Comparison of the dose rates of the non-neutron component (low LET) and the neutron one (high LET) of the radiation field at the aircraft flight level showed that the neutron component carried about 50% of the total dose. The dose rate measurements on the flights from the Middle Europe to the South and Middle America, then to Korea and Japan, showed that the flights over or near the equator region carried less dose rate; this was in accordance with the known geomagnetic latitude effect.

  1. Monitor units are not predictive of neutron dose for high-energy IMRT

    PubMed Central

    2012-01-01

    Background Due to the substantial increase in beam-on time of high energy intensity-modulated radiotherapy (>10 MV) techniques to deliver the same target dose compared to conventional treatment techniques, an increased dose of scatter radiation, including neutrons, is delivered to the patient. As a consequence, an increase in second malignancies may be expected in the future with the application of intensity-modulated radiotherapy. It is commonly assumed that the neutron dose equivalent scales with the number of monitor units. Methods Measurements of neutron dose equivalent were performed for an open and an intensity-modulated field at four positions: inside and outside of the treatment field at 0.2 cm and 15 cm depth, respectively. Results It was shown that the neutron dose equivalent, which a patient receives during an intensity-modulated radiotherapy treatment, does not scale with the ratio of applied monitor units relative to an open field irradiation. Outside the treatment volume at larger depth 35% less neutron dose equivalent is delivered than expected. Conclusions The predicted increase of second cancer induction rates from intensity-modulated treatment techniques can be overestimated when the neutron dose is simply scaled with monitor units. PMID:22883384

  2. Criticality prompt gamma and neutron dose equations validated by Monte Carlo analyses and compared to known criticality accident doses

    NASA Astrophysics Data System (ADS)

    Hochhalter, Eugene

    The United States (US) Department of Energy [DOE] and the Nuclear Regulatory Commission [NRC] have provided the nuclear industry with requirements, goals, and objectives for the preparation of safety analysis and the finalization of that safety analysis in the form of a documented safety analysis (DSA) and technical safety requirements (TSRs). The deterministic guidance provided by the NRC in Regulatory Guide (RG) 3.33 for calculating the prompt gamma and neutron doses from a criticality has a number of potential issues associated with the semi-empirical equations, which make these equations potentially out dated. The NRC guidance for estimating the prompt gamma and neutron doses to a facility worker due to an accidental criticality was withdrawn without newer deterministic guidance being issued. This research project determined the original basis for the RG prompt gamma and neutron equations, evaluated the potential issues associated with the RG 3.33 prompt gamma and neutron equations, and modified the RG 3.33 point source prompt gamma and neutron equations to calculate the doses for the selected set of criticality accidents. The criticality accidents addressed by this dissertation include: 1. U-235, Pu-239, and Pu-241 point source criticality, 2. U-235, Pu-239, and Pu-241 sphere source criticality, 3. Uranyl nitrate and plutonium nitrate solutions in a cylindrical process vessel and 4. Low level waste in 55-gallon and 30-gallon drums. The prompt gamma and neutron equation doses (RG 3.33/3.34/3.35) are compared to actual nuclear industry criticality accident worker doses to assess the conservatism of the RG equations. Finally, the RG 3.33 prompt gamma and neutron dose equations are compared to MCNP5 results to investigate consistency with respect to the modified prompt gamma and neutron dose equations and the representative dose estimates for each of the criticality configurations (point source, spherical source, and cylindrical source). Knowledge and accurate

  3. Absorbed dose rates in tissue from prompt gamma emissions from near-thermal neutron absorption

    DOE PAGES

    Schwahn, Scott O.

    2015-10-01

    Prompt gamma emission data from the International Atomic Energy Agency s Prompt Gamma-ray Neutron Activation Analysis database are analyzed to determine the absorbed dose rates in tissue to be expected when natural elements are exposed in a near-thermal neutron environment.

  4. Nominal effective radiation doses delivered during clinical trials of boron neutron capture therapy

    SciTech Connect

    Capala, J.; Diaz, A.Z.; Chanana, A.D.

    1997-12-31

    Boron neutron capture therapy (BNCT) is a binary system that, in theory, should selectively deliver lethal, high linear energy transfer (LET) radiation to tumor cells dispersed within normal tissues. It is based on the nuclear reaction 10-B(n, {alpha})7-Li, which occurs when the stable nucleus of boron-10 captures a thermal neutron. Due to the relatively high cross-section of the 10-B nucleus for thermal neutron capture and short ranges of the products of this reaction, tumor cells in the volume exposed to thermal neutrons and containing sufficiently high concentration of 10-B would receive a much higher radiation dose than the normal cells contained within the exposed volume. Nevertheless, radiation dose deposited in normal tissue by gamma and fast neutron contamination of the neutron beam, as well as neutron capture in nitrogen, 14-N(n,p)14-C, hydrogen, 1-H(n,{gamma})2-H, and in boron present in blood and normal cells, limits the dose that can be delivered to tumor cells. It is, therefore, imperative for the success of the BNCT the dosed delivered to normal tissues be accurately determined in order to optimize the irradiation geometry and to limit the volume of normal tissue exposed to thermal neutrons. These are the major objectives of BNCT treatment planning.

  5. Monte Carlo calculation of skyshine'' neutron dose from ALS (Advanced Light Source)

    SciTech Connect

    Moin-Vasiri, M.

    1990-06-01

    This report discusses the following topics on skyshine'' neutron dose from ALS: Sources of radiation; ALS modeling for skyshine calculations; MORSE Monte-Carlo; Implementation of MORSE; Results of skyshine calculations from storage ring; and Comparison of MORSE shielding calculations.

  6. Measurement of the neutron spectrum and ambient neutron dose rate equivalent from the small 252Cf source at 1 meter

    SciTech Connect

    Radev, R.

    2015-07-07

    NASA Langley Research Center requested a measurement of the neutron spectral distribution and fluence from the 252Cf source (model NS-120, LLNL serial # 7001677, referred as the SMALL Cf source) and determination of the ambient neutron dose rate equivalent and kerma at 100 cm for the Radiation Budget Instrument Experiment (Rad-X). The dosimetric quantities should be based on the neutron spectrum and the current neutron-to-dose conversion coefficients.

  7. Measuring the absolute deuterium-tritium neutron yield using the magnetic recoil spectrometer at OMEGA and the NIF.

    PubMed

    Casey, D T; Frenje, J A; Gatu Johnson, M; Séguin, F H; Li, C K; Petrasso, R D; Glebov, V Yu; Katz, J; Knauer, J P; Meyerhofer, D D; Sangster, T C; Bionta, R M; Bleuel, D L; Döppner, T; Glenzer, S; Hartouni, E; Hatchett, S P; Le Pape, S; Ma, T; MacKinnon, A; McKernan, M A; Moran, M; Moses, E; Park, H-S; Ralph, J; Remington, B A; Smalyuk, V; Yeamans, C B; Kline, J; Kyrala, G; Chandler, G A; Leeper, R J; Ruiz, C L; Cooper, G W; Nelson, A J; Fletcher, K; Kilkenny, J; Farrell, M; Jasion, D; Paguio, R

    2012-10-01

    A magnetic recoil spectrometer (MRS) has been installed and extensively used on OMEGA and the National Ignition Facility (NIF) for measurements of the absolute neutron spectrum from inertial confinement fusion implosions. From the neutron spectrum measured with the MRS, many critical implosion parameters are determined including the primary DT neutron yield, the ion temperature, and the down-scattered neutron yield. As the MRS detection efficiency is determined from first principles, the absolute DT neutron yield is obtained without cross-calibration to other techniques. The MRS primary DT neutron measurements at OMEGA and the NIF are shown to be in excellent agreement with previously established yield diagnostics on OMEGA, and with the newly commissioned nuclear activation diagnostics on the NIF.

  8. Determination the total neutron yields of several semiconductor compounds using various alpha emitters

    NASA Astrophysics Data System (ADS)

    Abdullah, Ramadhan Hayder; Sabr, Barzan Nehmat

    2016-03-01

    In the present work, the cross-sections of (α,n) reactions available in the literature as a function of α-particle energies for light and medium elements have been rearranged for α-particle energies from near threshold up to 10 MeV in steps of (0.050MeV) using the (Excel and Matlab) computer programs. The obtained data were used to calculate the neutron yields (n/106α) using the quick basic-computer program (Simpson Rules). The stopping powers of alpha particle energies from near threshold to 10 MeV for light and medium elements such as (nat.Be,10B,11B,13C,14N,nat.O,nat.F,nat.Mg,nat.Al,29Si,30Si, nat.P and 46.48Ti) have been calculated using the Zeigler formula. The kinetic energies (Tα) and the branching ratios of each α-emitters such as (211Bi, 210Po, 211Po, 215Po, 217At, 218Rn, 219Rn, 222Rn, 224Ra, 226Ra, 215Th, 228Th, 232U, 234U, 236U, 238U, 238Pu, 239Pu, 241Am, 245Es, 252Fm, 254Fm, 256Fm, 257Fm and 257Md) are taken into consideration to calculate the mean kinetic energy . The polynomial expressions were used to fitting the calculated weighted average of neutron yields (n/106α) for natural light and medium elements such as (Be, B, C, N, O, F, Mg, Al, Si, P and Ti) to determine the adopted neutron yields from the best fitting equation with minimum (CHISQ) at mean kinetic energies of various α-emitters. The total neutron yields (n/s/gx/ppmi) of the mentioned natural light and medium elements have been calculated using the adopted neutron yields (n/106α) from the fitting equations at mean kinetic energies of various α-emitters. The total neutron yields (n/s/gα-emitters/gcompounds) of semiconductor compounds such as (AlN, AlP, BN, BP, SiC, TiO2, BeSiN2, MgCN2, MgSiN2 and MgSiP2) have been calculated by mixing (1gram) of compounds with (1gram) of pure α-emitters using the quick basic computer program. The aim of the present work is to constructed and fabricate the neutron sources theoretically

  9. Analytical estimation of neutron yield in a micro gas-puff X pinch

    SciTech Connect

    Derzon, M. S.; Galambos, P. C.; Hagen, E. C.

    2012-12-01

    In this paper, we present the basic concepts for developing a micro x pinch as a small-scale neutron source. For compact sources, these concepts offer repetitive function at higher yields and pulsing rates than competing methods. The uniqueness of these concepts arises from the use of microelectronic technology to reduce the size of the target plasma and to efficiently heat the target gas. The use of repetitive microelectromechanical systems (MEMs) gas puff technology, as compared to cryogenic wires or solid targets (for the beam-target alternatives), has the potential to be robust and have a long lifetime because the plasma is not created from solid surfaces. The modeling suggests that a 50 J at the wall plug pulse could provide >10{sup 5} tritium (DT) neutrons and 10{sup 3} deuterium (DD) neutrons at temperatures of a few keV. At 1 kHz, this would be >10{sup 8} and 10{sup 6} neutrons per second, DT and DD, respectively, with a 250 {mu}m anode-cathode gap. DT gas puff devices may provide >10{sup 12} neutrons/s operating at 1 kHz and requiring 100 kW. The MEMs approach offers potentially high pulse rates and yields.

  10. Fusion neutron yield from a laser-irradiated heavy-water spray

    SciTech Connect

    Ter-Avetisyan, S.; Schnuerer, M.; Hilscher, D.; Jahnke, U.; Busch, S.; Nickles, P.V.; Sandner, W.

    2005-01-01

    The fusion neutron yield from a laser-irradiated heavy-water (D{sub 2}O) spray target was studied. Heavy-water droplets of about 150 nm diameter in the spray were exposed to 35 fs laser pulses at an intensity of 1x10{sup 19} W/cm{sup 2}. Due to the 10-50 times bigger size of the spray droplets compared to usual cluster sizes, deuterons are accelerated to considerably higher kinetic energies of up to 1 MeV. Neutrons are generated by the deuterons escaping from the plasma and initiating a fusion reaction within the surrounding cold plume of the spray jet. For each 0.6 J of laser pulse energy, 6x10{sup 3} neutrons are produced by about 10{sup 11} accelerated deuterons. This corresponds to a D(d,n) reaction probability of about 6x10{sup -8}. Compared to cluster targets, the reaction probability in the spray target is found to be two orders of magnitude larger. This finding apparently is due to both the considerably higher deuteron energies and the larger effective target thickness in the spray target. The measured neutron yield per accelerated deuteron [i.e., the D(d,n) reaction probability], is employed to compare and extrapolate the neutron emission characteristics from different target arrangements.

  11. Fission yields of In isotopes in the thermal neutron fission of235U

    NASA Astrophysics Data System (ADS)

    Shmid, M.; Engler, G.

    1983-03-01

    Fission yields of124 132In in the thermal neutron fission of235U were determined for the first time. Charge displacements ΔZ= Z p- Z UCD were calculated for the corresponding mass chains. Both fission yields and charge displacement values were compared with those obtained by systematics by Wahl et al. and Wolfsberg. It was found that the fission yields of the In isotopes obey the gaussian distribution. The displacement function of Wolfsberg seems to give the better representation of the experimental results. Half-lives of124 131In were determined from beta decay curves.

  12. Effect of wall thickness on measurement of dose for high energy neutrons.

    PubMed

    Perez-Nunez, Delia; Braby, Leslie A

    2010-01-01

    Neutrons produced from the interaction between galactic cosmic rays and spacecraft materials are responsible for a very important portion of the dose received by astronauts. The neutron energy spectrum depends on the incident charged particle spectrum and the scattering environment but generally extends to beyond 100 MeV. Tissue-equivalent proportional counters (TEPC) are used to measure the dose during the space mission, but their weight and size are very important factors for their design and construction. To achieve ideal neutron dosimetry, the wall thickness should be at least the range of a proton having the maximum energy of the neutrons to be monitored. This proton range is 0.1 cm for 10 MeV neutrons and 7.6 cm for 100 MeV neutrons. A 7.6 cm wall thickness TEPC would provide charged particle equilibrium (CPE) for neutrons up to 100 MeV, but for space applications it would not be reasonable in terms of weight and size. In order to estimate the errors in measured dose due to absence of CPE, MCNPX simulations of energy deposited by 10 MeV and 100 MeV neutrons in sites with wall thickness between 0.1 cm and 8.5 cm were performed. The results for 100 MeV neutrons show that energy deposition per incident neutron approaches a plateau as the wall thickness approaches 7.6 cm. For the 10 MeV neutrons, energy deposition per incident neutron decreases as the wall thickness increases above 0.1 cm due to attenuation. PMID:19959949

  13. Effect of wall thickness on measurement of dose for high energy neutrons.

    PubMed

    Perez-Nunez, Delia; Braby, Leslie A

    2010-01-01

    Neutrons produced from the interaction between galactic cosmic rays and spacecraft materials are responsible for a very important portion of the dose received by astronauts. The neutron energy spectrum depends on the incident charged particle spectrum and the scattering environment but generally extends to beyond 100 MeV. Tissue-equivalent proportional counters (TEPC) are used to measure the dose during the space mission, but their weight and size are very important factors for their design and construction. To achieve ideal neutron dosimetry, the wall thickness should be at least the range of a proton having the maximum energy of the neutrons to be monitored. This proton range is 0.1 cm for 10 MeV neutrons and 7.6 cm for 100 MeV neutrons. A 7.6 cm wall thickness TEPC would provide charged particle equilibrium (CPE) for neutrons up to 100 MeV, but for space applications it would not be reasonable in terms of weight and size. In order to estimate the errors in measured dose due to absence of CPE, MCNPX simulations of energy deposited by 10 MeV and 100 MeV neutrons in sites with wall thickness between 0.1 cm and 8.5 cm were performed. The results for 100 MeV neutrons show that energy deposition per incident neutron approaches a plateau as the wall thickness approaches 7.6 cm. For the 10 MeV neutrons, energy deposition per incident neutron decreases as the wall thickness increases above 0.1 cm due to attenuation.

  14. Dose dependence of the production yield of endohedral 133Xe-fullerene by ion implantation

    NASA Astrophysics Data System (ADS)

    Watanabe, S.; Ishioka, N. S.; Shimomura, H.; Muramatsu, H.; Sekine, T.

    2003-05-01

    The production yield of endohedral 133Xe-fullerene by ion implantation has been studied by taking advantage of the radioactivity of 133Xe. Fullerene targets, which were produced by vacuum evaporation of C 60 or C 70 on a Ni backing, were bombarded with 30-38 keV 133Xe ions by using an isotope separator at doses ranging from 1 × 10 12 to 1 × 10 14 cm -2. The production yield of endohedral 133Xe-fullerene was determined by an high performance liquid chromatography analysis following the dissolution of the targets in o-dichlorobenzene. It was found that the production yield decreased with increasing dose and incident energy, and the production yield of 133Xe@C 70 was higher than that of 133Xe@C 60 for the same dose and incident energy. Those production yields are discussed in connection with amorphization of fullerene molecules in collisions with 133Xe ions.

  15. Fractionated dose of 35-MEV fast neutrons and hypoxic tumor cell survival curve.

    PubMed

    U, R; Evans, J C; Cavanaugh, P J; Abramson, N; Thompson, T T; Wheless, D M

    1975-06-01

    The determination of the RBE for the MANTA fast neutrons produced by NRL is inprogress, with the model system using tumor cell population kinetic response patterns assayed in vitro after irradiation in vivo. Ascites tumor cells BW-5147 were irradiated with a clinically usable fast neutron beam from the NRL cyclotron, which is produced by accelrating deuterons to 35 MeV and using htem to bombard a thick berylliumtarget. The comparison of dose-effect relationships was made for doses ranging from30 to 1000 rads. The doses required for an isoeffect on BW-5147 hypoxic tumor cell survival and impairment of its reproductive capacity from fast neutron exposure were not different wheither it was given a single dose or the same dose given in three fractions separated by long recovery periods in situ. No intracellular repair of sublethal injury when the dose was given in three fractions, although the hypoxic BWp5147 tumor cells haveno effective reoxygenation or repopulation in this time interval. The RBE for the fast neutron beam is 4 relative to x rays for fractionated doses at the surviving fractionlevel of 0.6-0.7, while the RBE IS 2.5 FOR SINGLE DOSES. However, at a surviving fraction of 0.1, the RBE is 1.9 for single and 2.8 for fractionated doses. Analysis of thedaily cell population rate or mitotic delay between the two types of radiations at a similiar level of survival.

  16. DOSE PROFILE MODELING OF IDAHO NATIONAL LABORATORY’S ACTIVE NEUTRON INTERROGATION TEST FACILITY

    SciTech Connect

    D. L. Chichester; E. H. Seabury; J. M. Zabriskie; J. Wharton; A. J. Caffrey

    2009-06-01

    A new research and development laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for DT fusion (14.1 MeV) neutron generators (2 x 108 neutrons per second), DD fusion (2.5 MeV) neutron generators (up to 2 x 106 neutrons per second), and 252Cf spontaneous fission neutron sources (6.7 x 107 neutrons per second, 30 micrograms). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for 252Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield wall and entrance maze and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.

  17. Calculation of Ambient (H*(10)) and Personal (Hp(10)) Dose Equivalent from a 252Cf Neutron Source

    SciTech Connect

    Traub, Richard J.

    2010-03-26

    The purpose of this calculation is to calculate the neutron dose factors for the Sr-Cf-3000 neutron source that is located in the 318 low scatter room (LSR). The dose factors were based on the dose conversion factors published in ICRP-21 Appendix 6, and the Ambient dose equivalent (H*(10)) and Personal dose equivalent (Hp(10)) dose factors published in ICRP Publication 74.

  18. The alanine detector in BNCT dosimetry: Dose response in thermal and epithermal neutron fields

    SciTech Connect

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

    2015-01-15

    Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a {sup 60}Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes FLUKA and MCNP. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen and Olsen alanine response model. Results: The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. Conclusions: The

  19. Neutron Doses to the Concrete Vessel and Tendons of a Magnox Reactor Using Retrospective Dosimetry

    NASA Astrophysics Data System (ADS)

    Allen, D. A.; Thornton, D. A.; Wright, G. A.; Bird, A. J.; Rycroft, S.

    2009-08-01

    This paper describes the assessment of neutron doses to the concrete pressure vessels and stressing tendons above the cores of the Wylfa nuclear power plant. Following the observation of unexpected levels of activation in a routinely removed tendon from the top cap gallery, it was thought prudent to assess neutron doses to the vessel and its tendons and to consider the effect of potential radiation damage to the vessel integrity. To achieve this, the opportunity was taken to perform neutron activation measurements on tendon samples and to compare them with the results of predictions using the Monte Carlo code MCBEND. The results were used to underwrite neutron dose predictions for the concrete vessel and earlier results for the standpipes in this region of the reactor.

  20. Neutron fluences and dose equivalents measured with passive detectors on LDEF

    NASA Technical Reports Server (NTRS)

    Frank, A. L.; Benton, E. V.; Armstrong, T. W.; Colborn, B. L.

    1996-01-01

    Neutron fluences were measured on LDEF in the low energy (< 1 MeV) and high energy (> 1 MeV) ranges. The low energy detectors used the 6Li(n,alpha)T reaction with Gd foil absorbers to separate thermal (< 0.2 eV) and resonance (0.2 eV-1 MeV) neutron response. High energy detectors contained sets of fission foils (181Ta, 209Bi, 232Th, 238U) with different neutron energy thresholds. The measured neutron fluences together with predicted spectral shapes were used to estimate neutron dose equivalents. The detectors were located in the A0015 and P0006 experiments at the west and Earth sides of LDEF under shielding varying from 1 to 19 g/cm2. Dose equivalent rates varied from 0.8 to 3.3 microSv/d for the low energy neutrons and from 160 to 390 microSv/d for the high energy neutrons. This compares with TLD measured absorbed dose rates in the range of 1000-3000 microGy/d near these locations and demonstrates that high energy neutrons contribute a significant fraction of the total dose equivalent in LEO. Comparisons between measurements and calculations were made for high energy neutrons based on fission fragment tracks generated by fission foils at different shielding depths. A simple 1-D slab geometry was used in the calculations. Agreement between measurements and calculations depended on both shielding depth and threshold energy of the fission foils. Differences increased as both shielding and threshold energy increased. The modeled proton/neutron spectra appeared deficient at high energies. A 3-D model of the experiments is needed to help resolve the differences.

  1. Measurements of the neutron dose equivalent for various radiation qualities, treatment machines and delivery techniques in radiation therapy

    NASA Astrophysics Data System (ADS)

    Hälg, R. A.; Besserer, J.; Boschung, M.; Mayer, S.; Lomax, A. J.; Schneider, U.

    2014-05-01

    In radiation therapy, high energy photon and proton beams cause the production of secondary neutrons. This leads to an unwanted dose contribution, which can be considerable for tissues outside of the target volume regarding the long term health of cancer patients. Due to the high biological effectiveness of neutrons in regards to cancer induction, small neutron doses can be important. This study quantified the neutron doses for different radiation therapy modalities. Most of the reports in the literature used neutron dose measurements free in air or on the surface of phantoms to estimate the amount of neutron dose to the patient. In this study, dose measurements were performed in terms of neutron dose equivalent inside an anthropomorphic phantom. The neutron dose equivalent was determined using track etch detectors as a function of the distance to the isocenter, as well as for radiation sensitive organs. The dose distributions were compared with respect to treatment techniques (3D-conformal, volumetric modulated arc therapy and intensity-modulated radiation therapy for photons; spot scanning and passive scattering for protons), therapy machines (Varian, Elekta and Siemens linear accelerators) and radiation quality (photons and protons). The neutron dose equivalent varied between 0.002 and 3 mSv per treatment gray over all measurements. Only small differences were found when comparing treatment techniques, but substantial differences were observed between the linear accelerator models. The neutron dose equivalent for proton therapy was higher than for photons in general and in particular for double-scattered protons. The overall neutron dose equivalent measured in this study was an order of magnitude lower than the stray dose of a treatment using 6 MV photons, suggesting that the contribution of the secondary neutron dose equivalent to the integral dose of a radiotherapy patient is small.

  2. A response function calculation for a dose-equivalent neutron dosimeter using superheated drops

    SciTech Connect

    Wang, C.K. )

    1991-01-01

    A neutron dosimeter using superheated drops in gel was invented by Apfel. The SDD-100 or BD-100, which uses Freon-12 (CF{sub 2}Cl{sub 2}) for the superheated drops, is most useful in neutron dosimetry because it was claimed that the neutron response function of such a dosimeter is nearly dose equivalent. An ideal dose-equivalent neutron dosimeter should be totally independent of the energies of incident neutrons. Lo and Apfel have performed calculations and experiments to study the neutron response functions for various types of superheated drops, including Freon-12. Both their calculational and the experimental results demonstrated the dose-equivalent-like response function for the Freon-12. The agreement between the calculational results and the experimental results is not satisfactory, however, especially for neutrons with energies < 100 keV. One important factor, which was not considered and may have contributed to the disagreement, is the neutron-slowing-down effect. That is, kilo-electron-volt neutrons, although not energetic enough to trigger bubbles in Freon-12, have a short mean-free-path (< 1 cm) and can easily slow down or thermalize in the gel matrix and then trigger bubbles in Freon-12 via a {sup 35}Cl(n,p){sup 35}S reaction. To consider the slowing-down effect in the dosimeter, a neutron transport calculation must be performed. This paper describes the set of Monte Carlo neutron transport calculations that were performed to calculate the response function for a bare SDD-100 surrounded with various thicknesses of polyethylene (CH{sub 2}).

  3. Determination of pure neutron radiolysis yields for use in chemical modeling of supercritical water

    NASA Astrophysics Data System (ADS)

    Edwards, Eric J.

    This work has determined pure neutron radical yields at elevated temperature and pressure up to supercritical conditions using a reactor core radiation. The data will be necessary to provides realistic conditions for material corrosion experiments for the supercritical water reactor (SCWR) through water chemistry modeling. The work has been performed at the University of Wisconsin Nuclear Reactor using an apparatus designed to transport supercritical water near the reactor core. Low LET yield data used in the experiment was provided by a similar project at the Notre Dame Radiation Lab. Radicals formed by radiolysis were measured through chemical scavenging reactions. The aqueous electron was measured by two methods, a reaction with N2O to produce molecular nitrogen and a reaction with SF6 to produce fluoride ions. The hydrogen radical was measured through a reaction with ethanol-D6 (CD3CD2OD) to form HD. Molecular hydrogen was measured directly. Gaseous products were measured with a mass spectrometer and ions were measured with an ion selective electrode. Radiation energy deposition was calibrated for neutron and gamma radiation separately with a neutron activation analysis and a radiolysis experiment. Pure neutron yields were calculated by subtracting gamma contribution using the calibrated gamma energy deposition and yield results from work at the Notre Dame Radiation Laboratory. Pure neutron yields have been experimentally determined for aqueous electrons from 25°C to 400°C at 248 bar and for the hydrogen radical from 25°C to 350°C at 248 bar, Isothermal data has been acquired for the aqueous electron at 380°C and 400°C as a function of density. Molecular hydrogen yields were measured as a function of temperature and pressure, although there was evidence that chemical reactions with the walls of the water tubing were creating molecular hydrogen in addition to that formed through radiolysis. Critical hydrogen concentration behavior was investigated but a

  4. Neutron temporal diagnostic for high-yield deuterium-tritium cryogenic implosions on OMEGA

    NASA Astrophysics Data System (ADS)

    Stoeckl, C.; Boni, R.; Ehrne, F.; Forrest, C. J.; Glebov, V. Yu.; Katz, J.; Lonobile, D. J.; Magoon, J.; Regan, S. P.; Shoup, M. J.; Sorce, A.; Sorce, C.; Sangster, T. C.; Weiner, D.

    2016-05-01

    A next-generation neutron temporal diagnostic (NTD) capable of recording high-quality data for the highest anticipated yield cryogenic deuterium-tritium (DT) implosion experiments was recently installed at the Omega Laser Facility. A high-quality measurement of the neutron production width is required to determine the hot-spot pressure achieved in inertial confinement fusion experiments—a key metric in assessing the quality of these implosions. The design of this NTD is based on a fast-rise-time plastic scintillator, which converts the neutron kinetic energy to 350- to 450-nm-wavelength light. The light from the scintillator inside the nose-cone assembly is relayed ˜16 m to a streak camera in a well-shielded location. An ˜200× reduction in neutron background was observed during the first high-yield DT cryogenic implosions compared to the current NTD installation on OMEGA. An impulse response of ˜40 ± 10 ps was measured in a dedicated experiment using hard x-rays from a planar target irradiated with a 10-ps short pulse from the OMEGA EP laser. The measured instrument response includes contributions from the scintillator rise time, optical relay, and streak camera.

  5. The Role of the Driver Circuit in the Neutron Yield of the Plasma Focus

    NASA Astrophysics Data System (ADS)

    Sears, Jason; Schmidt, Andrea; Link, Anthony; Welch, Dale

    2015-11-01

    Emperical observations have suggested that dense plasma focus (DPF) neutron yield increases with driver impedance. Using the particle-in-cell code LSP, we reproduce this trend in a kJ DPF, and demonstrate in detail how driver impedance is coupled to neutron output. We implement a 2-D model of the plasma focus including self-consistent circuit-driven boundary conditions. We show that m=0 growth is central to beam formation and is a chaotic, non-deterministic process. Neutrons are produced when high, short-lived electric fields in the low-density cavity of an m=0 mode accelerate a beam of ions into the dense downstream pinch region. Neutron yield is highest when the ion beam is generated within 50 ns of the pinch formation on axis, because at that time the pinch (target) density is highest. High driver impedance contributes to prompt beam formation in two ways. First, the high impedance driver, losing less energy to run-down, has a faster run-in velocity and hence larger Rayleigh-Taylor features that more readily seed the m=0 instability. Second, the shorter anode of the high-impedance driver retains less trailing mass in the run-down region and thus exhibits fewer and less parasitic restrikes. Prepared by LLNL under Contract DE-AC52-07NA27344.

  6. Neutron temporal diagnostic for high-yield deuterium-tritium cryogenic implosions on OMEGA

    DOE PAGES

    Stoeckl, C.; Boni, R.; Ehrne, F.; Forrest, C. J.; Glebov, V. Yu.; Katz, J.; Lonobile, D. J.; Magoon, J.; Regan, S. P.; Shoup, III, M. J.; et al

    2016-05-10

    A next-generation neutron temporal diagnostic (NTD) capable of recording high-quality data for the highest anticipated yield cryogenic DT implosion experiments was recently installed at the Omega Laser Facility. A high-quality measurement of the neutron production width is required to determine the hot-spot pressure achieved in inertial confinement fusion experiments—a key metric in assessing the quality of these implosions. The design of this NTD is based on a fast-rise-time plastic scintillator, which converts the neutron kinetic energy to 350- to 450-nm-wavelength light. The light from the scintillator inside the nose-cone assembly is relayed ~16 m to a streak camera in amore » well-shielded location. An ~200× reduction in neutron background was observed during the first high-yield DT cryogenic implosions compared to the current NTD installation on OMEGA. An impulse response of ~40±10 ps was measured in a dedicated experiment using hard x rays from a planar target irradiated with a 10-ps short pulse from the OMEGA EP laser. Furthermore, the measured instrument response includes contributions from the scintillator rise time, optical relay, and streak camera.« less

  7. The calculation of neutron capture gamma-ray yields for space shielding applications

    NASA Technical Reports Server (NTRS)

    Yost, K. J.

    1972-01-01

    The application of nuclear models to the calculation of neutron capture and inelastic scattering gamma yields is discussed. The gamma ray cascade model describes the cascade process in terms of parameters which either: (1) embody statistical assumptions regarding electric and magnetic multipole transition strengths, level densities, and spin and parity distributions or (2) are fixed by experiment such as measured energies, spin and parity values, and transition probabilities for low lying states.

  8. Effect of Driver Impedance on Dense Plasma Focus Z-Pinch Neutron Yield and Beam Acceleration

    NASA Astrophysics Data System (ADS)

    Sears, J.; Link, A.; Ellsworth, J.; Falabella, S.; Rusnak, B.; Tang, V.; Schmidt, A.; Welch, D.

    2014-10-01

    We explore the effect of driver characteristics on dense plasma focus (DPF) neutron yield and beam acceleration using particle-in-cell (PIC) simulations of a kJ-scale DPF. Our PIC simulations are fluid for the run-down phase and transition to fully kinetic for the pinch phase. The anode-cathode boundary is driven by a circuit model of the capacitive driver, including system inductance, the load of the railgap switches, the guard resistors, and the coaxial transmission line parameters. Simulations are benchmarked to measurements of a table top kJ DPF experiment with neutron yield measured with He3-based detectors. Simulated neutron yield scales approximately with the fourth power of peak current, I4. We also probe the accelerating fields by measuring the acceleration of a 4 MeV deuteron beam and by measuring the DPF self-generated beam energy distribution, finding gradients higher than 50 MV/m. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344 and supported by the Laboratory Directed Research and Development Program (11-ERD-063) at LLNL.

  9. Analytic model to estimate thermonuclear neutron yield in z-pinches using the magnetic Noh problem

    NASA Astrophysics Data System (ADS)

    Allen, Robert C.

    The objective was to build a model which could be used to estimate neutron yield in pulsed z-pinch experiments, benchmark future z-pinch simulation tools and to assist scaling for breakeven systems. To accomplish this, a recent solution to the magnetic Noh problem was utilized which incorporates a self-similar solution with cylindrical symmetry and azimuthal magnetic field (Velikovich, 2012). The self-similar solution provides the conditions needed to calculate the time dependent implosion dynamics from which batch burn is assumed and used to calculate neutron yield. The solution to the model is presented. The ion densities and time scales fix the initial mass and implosion velocity, providing estimates of the experimental results given specific initial conditions. Agreement is shown with experimental data (Coverdale, 2007). A parameter sweep was done to find the neutron yield, implosion velocity and gain for a range of densities and time scales for DD reactions and a curve fit was done to predict the scaling as a function of preshock conditions.

  10. Study of the ratio of non-neutron to neutron dose components of cosmic radiation at typical commercial flight altitudes.

    PubMed

    Romero, A M; Saez-Vergara, J C; Rodriguez, R; Domínguez-Mompell, R

    2004-01-01

    CIEMAT, in close co-operation with Iberia Airlines, carried out an extensive programme of in-flight measurements, covering both hemispheres, during the years 2001 and 2002. Although the instrumentation onboard included different active devices, the results presented here were obtained from a polyethylene/tungsten-moderated rem meter (SWENDI2; Eberline) and an ionisation chamber (RSS-131; Reuter-Stokes) used for measuring the ambient dose equivalent due to the neutron and the non-neutron components of cosmic radiation, respectively. This paper presents a study of each of the dose components mentioned as a function of the vertical cut-off rigidity and the flight altitude. The ratio between the two components is also presented to determine the variations in cosmic radiation composition as a function of the aforementioned parameters. The experimental results have also been compared with those predicted by the code EPCARD3.2 for the non-neutron and the neutron components of the ambient dose equivalent.

  11. A new online detector for estimation of peripheral neutron equivalent dose in organ

    SciTech Connect

    Irazola, L. Sanchez-Doblado, F.; Lorenzoli, M.; Pola, A.; Bedogni, R.; Terrón, J. A.; Sanchez-Nieto, B.; Expósito, M. R.; Lagares, J. I.; Sansaloni, F.

    2014-11-01

    Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

  12. Neutron scattered dose equivalent to a fetus from proton radiotherapy of the mother.

    PubMed

    Mesoloras, Geraldine; Sandison, George A; Stewart, Robert D; Farr, Jonathan B; Hsi, Wen C

    2006-07-01

    Scattered neutron dose equivalent to a representative point for a fetus is evaluated in an anthropomorphic phantom of the mother undergoing proton radiotherapy. The effect on scattered neutron dose equivalent to the fetus of changing the incident proton beam energy, aperture size, beam location, and air gap between the beam delivery snout and skin was studied for both a small field snout and a large field snout. Measurements of the fetus scattered neutron dose equivalent were made by placing a neutron bubble detector 10 cm below the umbilicus of an anthropomorphic Rando phantom enhanced by a wax bolus to simulate a second trimester pregnancy. The neutron dose equivalent in milliSieverts (mSv) per proton treatment Gray increased with incident proton energy and decreased with aperture size, distance of the fetus representative point from the field edge, and increasing air gap. Neutron dose equivalent to the fetus varied from 0.025 to 0.450 mSv per proton Gray for the small field snout and from 0.097 to 0.871 mSv per proton Gray for the large field snout. There is likely to be no excess risk to the fetus of severe mental retardation for a typical proton treatment of 80 Gray to the mother since the scattered neutron dose to the fetus of 69.7 mSv is well below the lower confidence limit for the threshold of 300 mGy observed for the occurrence of severe mental retardation in prenatally exposed Japanese atomic bomb survivors. However, based on the linear no threshold hypothesis, and this same typical treatment for the mother, the excess risk to the fetus of radiation induced cancer death in the first 10 years of life is 17.4 per 10,000 children. PMID:16898451

  13. Neutron scattered dose equivalent to a fetus from proton radiotherapy of the mother.

    PubMed

    Mesoloras, Geraldine; Sandison, George A; Stewart, Robert D; Farr, Jonathan B; Hsi, Wen C

    2006-07-01

    Scattered neutron dose equivalent to a representative point for a fetus is evaluated in an anthropomorphic phantom of the mother undergoing proton radiotherapy. The effect on scattered neutron dose equivalent to the fetus of changing the incident proton beam energy, aperture size, beam location, and air gap between the beam delivery snout and skin was studied for both a small field snout and a large field snout. Measurements of the fetus scattered neutron dose equivalent were made by placing a neutron bubble detector 10 cm below the umbilicus of an anthropomorphic Rando phantom enhanced by a wax bolus to simulate a second trimester pregnancy. The neutron dose equivalent in milliSieverts (mSv) per proton treatment Gray increased with incident proton energy and decreased with aperture size, distance of the fetus representative point from the field edge, and increasing air gap. Neutron dose equivalent to the fetus varied from 0.025 to 0.450 mSv per proton Gray for the small field snout and from 0.097 to 0.871 mSv per proton Gray for the large field snout. There is likely to be no excess risk to the fetus of severe mental retardation for a typical proton treatment of 80 Gray to the mother since the scattered neutron dose to the fetus of 69.7 mSv is well below the lower confidence limit for the threshold of 300 mGy observed for the occurrence of severe mental retardation in prenatally exposed Japanese atomic bomb survivors. However, based on the linear no threshold hypothesis, and this same typical treatment for the mother, the excess risk to the fetus of radiation induced cancer death in the first 10 years of life is 17.4 per 10,000 children.

  14. Neutron scattered dose equivalent to a fetus from proton radiotherapy of the mother

    SciTech Connect

    Mesoloras, Geraldine; Sandison, George A.; Stewart, Robert D.; Farr, Jonathan B.; Hsi, Wen C.

    2006-07-15

    Scattered neutron dose equivalent to a representative point for a fetus is evaluated in an anthropomorphic phantom of the mother undergoing proton radiotherapy. The effect on scattered neutron dose equivalent to the fetus of changing the incident proton beam energy, aperture size, beam location, and air gap between the beam delivery snout and skin was studied for both a small field snout and a large field snout. Measurements of the fetus scattered neutron dose equivalent were made by placing a neutron bubble detector 10 cm below the umbilicus of an anthropomorphic Rando[reg] phantom enhanced by a wax bolus to simulate a second trimester pregnancy. The neutron dose equivalent in milliSieverts (mSv) per proton treatment Gray increased with incident proton energy and decreased with aperture size, distance of the fetus representative point from the field edge, and increasing air gap. Neutron dose equivalent to the fetus varied from 0.025 to 0.450 mSv per proton Gray for the small field snout and from 0.097 to 0.871 mSv per proton Gray for the large field snout. There is likely to be no excess risk to the fetus of severe mental retardation for a typical proton treatment of 80 Gray to the mother since the scattered neutron dose to the fetus of 69.7 mSv is well below the lower confidence limit for the threshold of 300 mGy observed for the occurrence of severe mental retardation in prenatally exposed Japanese atomic bomb survivors. However, based on the linear no threshold hypothesis, and this same typical treatment for the mother, the excess risk to the fetus of radiation induced cancer death in the first 10 years of life is 17.4 per 10 000 children.

  15. Determination of canine dose conversion factors in mixed neutron and gamma radiation fields. Technical report

    SciTech Connect

    Torres, B.A.; Bhatt, R.C.; Myska, J.C.; Holland, B.K.

    1996-07-01

    The primary objective of mixed-field neutron/gamma radiation dosimetry in canine irradiation experiments conducted at the Armed Forces Radiobiology Research Institute (AFRRI) is to determine the absorbed midline tissue dose (MLT) at the region of interest in the canine. A dose conversion factor (DCF) can be applied to free-in-air (FIA) dose measurements to estimate the MLT doses to canines. This report is a summary of the measured DCFs that were used to determine the MLT doses in canines at AFRRI from 1979 to 1992.

  16. Improvements in Fabrication of Elastic Scattering Foils Used to Measure Neutron Yield by the Magnetic Recoil Spectrometer

    DOE PAGES

    Reynolds, H. G.; Schoff, M. E.; Farrell, M. P.; Gatu Johnson, M.; Bionta, R. M.; Frenje, J. A.

    2016-08-01

    The magnetic recoil spectrometer uses a deuterated polyethylene polymer (CD2) foil to measure neutron yield in inertial confinement fusion experiments. Higher neutron yields in recent experiments have resulted in primary signal saturation in the detector CR-39 foils, necessitating the fabrication of thinner CD2 foils than established methods could provide. A novel method of fabricating deuterated polymer foils is described. The resulting foils are thinner, smoother, and more uniform in thickness than the foils produced by previous methods. Here, these new foils have successfully been deployed at the National Ignition Facility, enabling higher neutron yield measurements than previous foils, with nomore » primary signal saturation.« less

  17. Calculation of effective dose from measurements of secondary neutron spectra and scattered photon dose from dynamic MLC IMRT for 6 MV, 15 MV, and 18 MV beam energies.

    PubMed

    Howell, Rebecca M; Hertel, Nolan E; Wang, Zhonglu; Hutchinson, Jesson; Fullerton, Gary D

    2006-02-01

    Effective doses were calculated from the delivery of 6 MV, 15 MV, and 18 MV conventional and intensity-modulated radiation therapy (IMRT) prostate treatment plans. ICRP-60 tissue weighting factors were used for the calculations. Photon doses were measured in phantom for all beam energies. Neutron spectra were measured for 15 MV and 18 MV and ICRP-74 quality conversion factors used to calculate ambient dose equivalents. The ambient dose equivalents were corrected for each tissue using neutron depth dose data from the literature. The depth corrected neutron doses were then used as a measure of the neutron component of the ICRP protection quantity, organ equivalent dose. IMRT resulted in an increased photon dose to many organs. However, the IMRT treatments resulted in an overall decrease in effective dose compared to conventional radiotherapy. This decrease correlates to the ability of an intensity-modulated field to minimize dose to critical normal structures in close proximity to the treatment volume. In a comparison of the three beam energies used for the IMRT treatments, 6 MV resulted in the lowest effective dose, while 18 MV resulted in the highest effective dose. This is attributed to the large neutron contribution for 18 MV compared to no neutron contribution for 6 MV. PMID:16532941

  18. Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am-Be neutron source.

    PubMed

    Li, Gang; Xu, Jiayun; Zhang, Jie

    2014-10-22

    Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am-Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am-Be neutron source was investigated by measuring the count rates obtained through a (3)He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, (3)He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter.

  19. Fission Fragment Yield, Cross Section and Prompt Neutron and Gamma Emission Data from Actinide Isotopes

    NASA Astrophysics Data System (ADS)

    Hambsch, F.-J.; Oberstedt, S.; Al-Adili, A.; Brys, T.; Billnert, R.; Matei, C.; Oberstedt, A.; Salvador-Castiñeira, P.; Tudora, A.; Vidali, M.

    2014-05-01

    Recent experimental investigations on major and minor actinides at the JRC-IRMM are presented. Fission-fragment distributions of isotopes with vibrational resonances in the sub-threshold fission cross section, i. e. 234,238U, have been measured. For 234U, the impact of an increased neutron multiplicity for the heavy fragments with higher incident neutron energies has been studied as observed in experiment and also recently theoretically predicted. The impact is found to be noticeable on post-neutron mass yields, which are the relevant quantities for a-priori waste assessments. The fission cross sections for 240,242Pu at threshold and in the plateau region are being investigated within the ANDES project. The results show some discrepancies to the ENDF/B-VII.1 evaluation mainly for 242Pu around 1 MeV, where the evaluation exhibits a resonance-like structure not observed so clearly in the present work. The requested target accuracy in design studies of innovative reactor concepts like Gen-IV is in the range of a few percent. In order to be able to respond to requests for measurements of prompt neutron and γ-ray emission in fission JRC-IRMM has also invested in setting up a neutron and γ-ray detector array. The neutron array is called SCINTIA and has so far been tested with 252Cf(SF). For γ-ray multiplicity and spectrum measurements of 252Cf(SF) and 235U(nth, f) lanthanum- and cerium-halide detectors were successfully used.

  20. Dose-equivalent response CR-39 track detector for personnel neutron dosimetry

    NASA Astrophysics Data System (ADS)

    Oda, K.; Ito, M.; Yoneda, H.; Miyake, H.; Yamamoto, J.; Tsuruta, T.

    1991-09-01

    A dose-equivalent response detector based on CR-39 has been designed to be applied for personnel neutron dosimetry. The intrinsic detection efficiency of bare CR-39 was first evaluated from irradiation experiments with monoenergetic neutrons and theoretical calculations. In the second step, the radiator effect was investigated for the purpose of sensitization to fast neutrons. A two-layer radiator consisting of deuterized dotriacontane (C 32D 66) and polyethylene (CH 2) was designed. Finally, we made the CR-39 detector sensitive to thermal neutrons by doping with orthocarborane (B 10H 122C 2), and also estimated the contribution of albedo neutrons. It was found that the new detector — boron-doped CR-39 with the two-layer radiator — would have a flat response with an error of about 70% in a wide energy region, ranging from thermal to 15 MeV.

  1. Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls.

    PubMed

    Kyllönen, J; Lindborg, L

    2007-01-01

    A graphite-walled proportional counter with low neutron sensitivity was used in combination with a tissue-equivalent proportional counter (TEPC) to separate the photon and neutron components in mixed radiation fields. Monte Carlo (MCNP4C) simulations of the photon and neutron responses of the two detectors were done to obtain correction factors for the sensitivity differences. In an alternative method the radiation components were determined using constant-yD-values for typical photon and neutron energy distributions. The results show no significant difference between the two methods and the measured neutron dose-equivalent agrees within +/-50% with Bonner sphere determined values. The experimental data were obtained in measurement campaigns organised within the EVIDOS-project. PMID:17309871

  2. An estimation of the yield and response functions for the mini neutron monitor

    NASA Astrophysics Data System (ADS)

    Caballero-Lopez, R. A.

    2016-08-01

    The present study estimates the yield and response functions of the mini neutron monitor (miniNM). This relatively new cosmic ray detector is the mobile version of the standard NM64. It can be use not only to calibrate the NM64 but also to study the modulation processes. Due to its portability, the miniNM can be easily placed in a suitable location to measure secondary particles, which give information about the intensity variations of galactic and solar cosmic rays. In order to perform these modulation studies with miniNMs, it is crucial to know their sensitivity to detect secondary cosmic ray flux, i.e., we must know their yield function. A previous study found that miniNM and NM64 have slightly different response functions. This work analyzes the observed counting rate ratio (miniNM to NM64) and gives for the first time an useful expression for the yield function of the miniNM. The results found here will allow to interpret the new measurements with this mobile neutron monitor. For comparison, a brief summary of the NM64 yield functions reported by other authors is presented.

  3. Spot scanning proton therapy minimizes neutron dose in the setting of radiation therapy administered during pregnancy.

    PubMed

    Wang, Xin; Poenisch, Falk; Sahoo, Narayan; Zhu, Ronald X; Lii, MingFwu; Gillin, Michael T; Li, Jing; Grosshans, David

    2016-01-01

    This is a real case study to minimize the neutron dose equivalent (H) to a fetus using spot scanning proton beams with favorable beam energies and angles. Minimum neutron dose exposure to the fetus was achieved with iterative planning under the guidance of neutron H measurement. Two highly conformal treatment plans, each with three spot scanning beams, were planned to treat a 25-year-old pregnant female with aggressive recurrent chordoma of the base of skull who elected not to proceed with termination. Each plan was scheduled for delivery every other day for robust target coverage. Neutron H to the fetus was measured using a REM500 neutron survey meter placed at the fetus position of a patient simulating phantom. 4.1 and 44.1 μSv/fraction were measured for the two initial plans. A vertex beam with higher energy and the fetal position closer to its central axis was the cause for the plan that produced an order higher neutron H. Replacing the vertex beam with a lateral beam reduced neutron H to be comparable with the other plan. For a prescription of 70 Gy in 35 fractions, the total neutron H to the fetus was estimated to be 0.35 mSv based on final measurement in single fraction. In comparison, the passive scattering proton plan and photon plan had an estimation of 26 and 70 mSv, respectively, for this case. While radiation therapy in pregnant patients should be avoided if at all possible, our work demonstrated spot scanning beam limited the total neutron H to the fetus an order lower than the suggested 5 mSv regulation threshold. It is far superior than passive scattering beam and careful beam selection with lower energy and keeping fetus further away from beam axis are essential in minimizing the fetus neutron exposure. PMID:27685136

  4. Elastic stability of high dose neutron irradiated spinel

    SciTech Connect

    Li, Z.; Chan, S.K.; Garner, F.A.

    1995-04-01

    The objective of this effort is to identify ceramic materials that are suitable for fusion reactor applications. Elastic constants (C{sub 11}, C{sub 12}, and C{sub 44}) of spinel (MgAl{sub 2}O{sub 4}) single crystals irradiated to very high neutron fluences have geen measured by an ultrasonic technique. Although results of a neutron diffraction study show that cation occupation sites are significantly changed in the irradiated samples, no measurable differences occurred in their elastic properties. In order to understand such behavior, the elastic properties of a variety of materials with either normal or inverse spinel structures were studied. The cation valence and cation distribution appear to have little influence on the elastic properties of spinel materials.

  5. Polar-Drive Designs for Optimizing Neutron Yields on the National Ignition Faciltiy

    SciTech Connect

    Cok, A.M.; Craxton, R.S.; McKenty, P.W.

    2008-09-10

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350 kJ to 1.5 MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 10^15–10^16 are expected.

  6. Polar-drive designs for optimizing neutron yields on the National Ignition Facility

    SciTech Connect

    Cok, A. M.; Craxton, R. S.; McKenty, P. W.

    2008-08-15

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350 kJ to 1.5 MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 10{sup 15}-10{sup 16} are expected.

  7. Polar-drive designs for optimizing neutron yields on the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Cok, A. M.; Craxton, R. S.; McKenty, P. W.

    2008-08-01

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350kJto1.5MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 1015-1016 are expected.

  8. Quantitative assessment of the cataractogenic potential of very low doses of neutrons

    NASA Technical Reports Server (NTRS)

    Worgul, B. V.; Medvedovsky, C.; Huang, Y.; Marino, S. A.; Randers-Pehrson, G.; Brenner, D. J.

    1996-01-01

    We report on the prevalence and relative biological effectiveness (RBE) for various stages of lens opacification in rats induced by very low doses (2 to 250 mGy) of medium-energy (440 keV) neutrons, compared to those for X rays. Neutron doses were delivered either in a single fraction or in four separate fractions and the irradiated animals were followed for over 100 weeks. At the highest observed dose (250 mGy) and at early observation times, there was evidence of an inverse dose-rate effect; i.e., a fractionated exposure was more potent than a single exposure. Neutron RBEs relative to X rays were estimated using a non-parametric technique. The results were only weakly dependent on time postirradiation. At 30 weeks, for example, 80% confidence intervals for the RBE of acutely delivered neutrons relative to X rays were 8-16 at 250 mGy, 10-20 at 50 mGy, 50-100 at 10 mGy and 250-500 at 2 mGy. The results are consistent with the estimated neutron RBEs in Japanese A-bomb survivors, though broad confidence bounds are present in the Japanese results. Our findings are also consistent with data reported earlier for cataractogenesis induced by heavy ions in rats, mice, and rabbits. We conclude from these results that, at very low doses (<10 mGy), the RBE for neutron-induced cataractogenesis is considerably larger than the RBE of 20 commonly used, and use of a significantly larger value for calculating equivalent dose would be prudent.

  9. Microdosimetric measurements for neutron-absorbed dose determination during proton therapy

    PubMed Central

    Pérez-Andújar, Angélica; DeLuca, Paul M.; Thornton, Allan F.; Fitzek, Markus; Hecksel, Draik; Farr, Jonathan

    2012-01-01

    This work presents microdosimetric measurements performed at the Midwest Proton Radiotherapy Institute in Bloomington, Indiana, USA. The measurements were done simulating clinical setups with a water phantom and for a variety of stopping targets. The water phantom was irradiated by a proton spread out Bragg peak (SOBP) and by a proton pencil beam. Stopping target measurements were performed only for the pencil beam. The targets used were made of polyethylene, brass and lead. The objective of this work was to determine the neutron-absorbed dose for a passive and active proton therapy delivery, and for the interactions of the proton beam with materials typically in the beam line of a proton therapy treatment nozzle. Neutron doses were found to be higher at 45° and 90° from the beam direction for the SOBP configuration by a factor of 1.1 and 1.3, respectively, compared with the pencil beam. Meanwhile, the pencil beam configuration produced neutron-absorbed doses 2.2 times higher at 0° than the SOBP. For stopping targets, lead was found to dominate the neutron-absorbed dose for most angles due to a large production of low-energy neutrons emitted isotropically. PMID:22334761

  10. Effect of cathode structure on neutron yield performance of a miniature plasma focus device

    NASA Astrophysics Data System (ADS)

    Verma, Rishi; Rawat, R. S.; Lee, P.; Lee, S.; Springham, S. V.; Tan, T. L.; Krishnan, M.

    2009-07-01

    In this Letter we report the effect of two different cathode structures - tubular and squirrel cage, on neutron output from a miniature plasma focus device. The squirrel cage cathode is typical of most DPF sources, with an outer, tubular envelope that serves as a vacuum housing, but does not carry current. The tubular cathode carries the return current and also serves as the vacuum envelope, thereby minimizing the size of the DPF head. The maximum average neutron yield of (1.82±0.52)×10 n/shot for the tubular cathode at 4 mbar was enhanced to (1.15±0.2)×10 n/shot with squirrel cage cathode at 6 mbar operation. These results are explained on the basis of a current sheath loading/mass choking effect. The penalty for using a non-transparent cathode negates the advantage of the smaller size of the DPF head.

  11. Dose estimation for internal organs during boron neutron capture therapy for body-trunk tumors.

    PubMed

    Sakurai, Y; Tanaka, H; Suzuki, M; Masunaga, S; Kinashi, Y; Kondo, N; Ono, K; Maruhashi, A

    2014-06-01

    Radiation doses during boron neutron capture therapy for body-trunk tumors were estimated for various internal organs, using data from patients treated at Kyoto University Research Reactor Institute. Dose-volume histograms were constructed for tissues of the lung, liver, kidney, pancreas, and bowel. For pleural mesothelioma, the target total dose to the normal lung tissues on the diseased side is 5Gy-Eq in average for the whole lung. It was confirmed that the dose to the liver should be carefully considered in cases of right lung disease.

  12. Hair 32P measurement for body dose mapping in non-fatal exposures to fast neutrons.

    PubMed

    Mianji, Fereidoun A; Jafari, Sheyda; Zaryouni, Saiedeh; Hajizadeh, Bardia

    2015-03-01

    Dosimetry bioassay methods are the backbone of a personal dosimetry in criticality accidents. Although methods like hair dosimetry and the use of activation foils (e.g., (32)S) have been employed for decades, capabilities of different techniques, effects of hair type and neutron spectrum on the dose response, sensitivity and uncertainties of different techniques, etc., need more investigations. For this reason, the use of the (32)S(n,p)(32)P reaction and hair samples for estimating non-fatal doses from fast neutrons was studied. The experiments were carried out with the hair samples attached on a RANDO phantom in a Cf-252 neutron field, in the dose range of about 0.05-1.15 Gy. In addition, the adequate post-accident preparation for hair samples including optimum conditioning and timing were investigated. Experimental results prove the good sensitivity and merit of the method for neutron quantification in the mentioned dose range for which other bioassay methods are of poor resolution and sensitivity. A rough estimation of the dose-response curve for Iranian hair was also derived.

  13. Hair 32P measurement for body dose mapping in non-fatal exposures to fast neutrons.

    PubMed

    Mianji, Fereidoun A; Jafari, Sheyda; Zaryouni, Saiedeh; Hajizadeh, Bardia

    2015-03-01

    Dosimetry bioassay methods are the backbone of a personal dosimetry in criticality accidents. Although methods like hair dosimetry and the use of activation foils (e.g., (32)S) have been employed for decades, capabilities of different techniques, effects of hair type and neutron spectrum on the dose response, sensitivity and uncertainties of different techniques, etc., need more investigations. For this reason, the use of the (32)S(n,p)(32)P reaction and hair samples for estimating non-fatal doses from fast neutrons was studied. The experiments were carried out with the hair samples attached on a RANDO phantom in a Cf-252 neutron field, in the dose range of about 0.05-1.15 Gy. In addition, the adequate post-accident preparation for hair samples including optimum conditioning and timing were investigated. Experimental results prove the good sensitivity and merit of the method for neutron quantification in the mentioned dose range for which other bioassay methods are of poor resolution and sensitivity. A rough estimation of the dose-response curve for Iranian hair was also derived. PMID:25503945

  14. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

  15. Dose-response relationship of dicentric chromosomes in human lymphocytes obtained for the fission neutron therapy facility MEDAPP at the research reactor FRM II.

    PubMed

    Schmid, E; Wagner, F M; Romm, H; Walsh, L; Roos, H

    2009-02-01

    The biological effectiveness of neutrons from the neutron therapy facility MEDAPP (mean neutron energy 1.9 MeV) at the new research reactor FRM II at Garching, Germany, has been analyzed, at different depths in a polyethylene phantom. Whole blood samples were exposed to the MEDAPP beam in special irradiation chambers to total doses of 0.14-3.52 Gy at 2-cm depth, and 0.18-3.04 Gy at 6-cm depth of the phantom. The neutron and gamma-ray absorbed dose rates were measured to be 0.55 Gy min(-1) and 0.27 Gy min(-1) at 2-cm depth, while they were 0.28 and 0.25 Gy min(-1) at 6-cm depth. Although the irradiation conditions at the MEDAPP beam and the RENT beam of the former FRM I research reactor were not identical, neutrons from both facilities gave a similar linear-quadratic dose-response relationship for dicentric chromosomes at a depth of 2 cm. Different dose-response curves for dicentrics were obtained for the MEDAPP beam at 2 and 6 cm depth, suggesting a significantly lower biological effectiveness of the radiation with increasing depth. No obvious differences in the dose-response curves for dicentric chromosomes estimated under interactive or additive prediction between neutrons or gamma-rays and the experimentally obtained dose-response curves could be determined. Relative to (60)Co gamma-rays, the values for the relative biological effectiveness at the MEDAPP beam decrease from 5.9 at 0.14 Gy to 1.6 at 3.52 Gy at 2-cm depth, and from 4.1 at 0.18 Gy to 1.5 at 3.04 Gy at 6-cm depth. Using the best possible conditions of consistency, i.e., using blood samples from the same donor and the same measurement techniques for about two decades, avoiding the inter-individual variations in sensitivity or the differences in methodology usually associated with inter-laboratory comparisons, a linear-quadratic dose-response relationship for the mixed neutron and gamma-ray MEDAPP field as well as for its fission neutron part was obtained. Therefore, the debate on whether the fission-neutron

  16. Oxidative Damage and Mutagenic Potency of Fast Neutron and UV-B Radiation in Pollen Mother Cells and Seed Yield of Vicia faba L.

    PubMed Central

    Abdel Haliem, Ekram; Abdullah, Hanan; AL-Huqail, Asma A.

    2013-01-01

    In recent years, there has been a great deal of attention toward free radicals, reactive oxygen species (ROS) generated by exposure of crop plant cells to physical radiations. Henceforth, the current study was planned to compare oxidative stress and mutagenic potential of different irradiation doses of fast neutron (FN) and UV-B on meiotic-pollen mother cells (PMCs), pollen grains (PGs) and seeds yielded from irradiated faba beans seedlings. On the cytogenetic level, each irradiation type had special interference with DNA of PMC and exhibited wide range of mutagenic action on the frequency and type of chromosomal anomalies, fertility of PGs and seed yield productivity based on the irradiation exposure dose and radiation sensitivity of faba bean plants compared with un-irradiated ones. On the molecular level, SDS-PAGE and RPAD-PCR analyses of seeds yielded from irradiated seedlings exhibited distinctive polymorphisms based on size, intensity, appearance, and disappearance of polypeptides bands compared with un-irradiated ones. The total values of protein and DNA polymorphisms reached 88% and 90.80% respectively. The neutron fluency (2.3 × 106 n/cm2) and UV-B dose for 1 hr were recorded as bio-positive effects. The present study proved that genetic variations revealed by cytogenetic test could be supported by gene expression (alterations in RAPD and protein profiles). PMID:24066298

  17. Boron Neutron Capture Therapy (BNCT) Dose Calculation using Geometrical Factors Spherical Interface for Glioblastoma Multiforme

    SciTech Connect

    Zasneda, Sabriani; Widita, Rena

    2010-06-22

    Boron Neutron Capture Therapy (BNCT) is a cancer therapy by utilizing thermal neutron to produce alpha particles and lithium nuclei. The superiority of BNCT is that the radiation effects could be limited only for the tumor cells. BNCT radiation dose depends on the distribution of boron in the tumor. Absorbed dose to the cells from the reaction 10B (n, {alpha}) 7Li was calculated near interface medium containing boron and boron-free region. The method considers the contribution of the alpha particle and recoiled lithium particle to the absorbed dose and the variation of Linear Energy Transfer (LET) charged particles energy. Geometrical factor data of boron distribution for the spherical surface is used to calculate the energy absorbed in the tumor cells, brain and scalp for case Glioblastoma Multiforme. The result shows that the optimal dose in tumor is obtained for boron concentrations of 22.1 mg {sup 10}B/g blood.

  18. Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358 C

    SciTech Connect

    Cockeram, Brian V; Smith, Richard W; Leonard, Keith J; Byun, Thak Sang; Snead, Lance Lewis

    2011-01-01

    Wrought Zircaloy-2 and Zircaloy-4 were neutron irradiated at nominally 358 C in the high flux isotope reactor (HFIR) at relatively low neutron fluences between 5.8 1022 and 2.9 1025 n/m2 (E > 1 MeV). The irradiation hardening and change in microstructure were characterized following irradiation using tensile testing and examinations of microstructure using Analytical Electron Microscopy (AEM). Small increments of dose (0.0058, 0.11, 0.55, 1.08, and 2.93 1025 n/m2) were used in the range where the saturation of irradiation hardening is typically observed so that the role of microstructure evolution and hai loop formation on irradiation hardening could be correlated. An incubation dose between 5.8 1023 and 1.1 1024 n/m2 was needed for loop nucleation to occur that resulted in irradiation hardening. Increases in yield strength were consistent with previous results in this temperature regime, and as expected less irradiation hardening and lower hai loop number density values than those generally reported in literature for irradiations at 260 326 C were observed. Unlike previous lower temperature data, there is evidence in this study that the irradiation hardening can decrease with dose over certain ranges of fluence. Irradiation induced voids were observed in very low numbers in the Zircaloy-2 materials at the highest fluence.

  19. Determination of carrier yields for neutron activation analysis using energy dispersive X-ray spectrometry

    USGS Publications Warehouse

    Johnson, R.G.; Wandless, G.A.

    1984-01-01

    A new method is described for determining carrier yield in the radiochemical neutron activation analysis of rare-earth elements in silicate rocks by group separation. The method involves the determination of the rare-earth elements present in the carrier by means of energy-dispersive X-ray fluorescence analysis, eliminating the need to re-irradiate samples in a nuclear reactor after the gamma ray analysis is complete. Results from the analysis of USGS standards AGV-1 and BCR-1 compare favorably with those obtained using the conventional method. ?? 1984 Akade??miai Kiado??.

  20. Neutron Yield and Ion Temperature from DD and DT Fusion in National Ignition Facility High-Foot Implosions

    NASA Astrophysics Data System (ADS)

    Knauer, J. P.; Gatu Johnson, M.; Frenje, J. A.; Petrasso, R. D.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Cerjan, C. J.; Doeppner, T.; Eckart, M. J.; Grim, G. P.; Hartouni, E. P.; Hatarik, R.; Hinkel, D. E.; Hurricane, O. A.; Kritcher, A.; Le Pape, S.; Ma, T.; Munro, D. H.; Patel, P. K.; Ralph, J. E.; Sayre, D. B.; Spears, B. K.; Yeamans, C. B.; Kilkenny, J. D.

    2015-11-01

    Simultaneous measures of neutrons emitted from DT fusion implosions are postulated to provide insight into the fuel conditions during neutron emission. Neutron spectral diagnostics of National Ignition Facility ``high-foot'' implosions measure both the DT and DD fusion neutron spectra. Equivalent ion temperature is measured from the width of the DT and DD neutron emission and the respective yields from the peak areas. This work has focused on reasons for differing inferred temperatures from the DT and DD spectra and the yield ratio. Spatial and temporal averages of the DT and DD reactivities as corrections to the homogeneous and static temperature distributions are shown. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  1. Neutron spectrum and yield of the Hiroshima A-bomb deduced from radionuclide measurements at one location.

    PubMed

    Rühm, W; Kato, K; Korschinek, G; Morinaga, H; Nolte, E

    1995-07-01

    In this paper measurements of the radionuclides of 36Cl, 41Ca, 60Co, 152Eu and 154Eu in samples from Hiroshima, which were exposed to neutrons of the A-bomb explosion, are interpreted. In order to calculate the neutron spectrum at the sample site, neutron transport calculations using Monte Carlo techniques were carried out. Activation profiles in a granite mock-up irradiated with reactor neutrons could be reproduced by this method using DS86 input parameters. The calculated neutron spectrum at the sample site for non-thermal neutrons is identical to that obtained in DS86, but contains some 50% more thermal neutrons. The influence of parameters like soil composition, source terms and air humidity on the activation of these radioisotopes is discussed. The granite-covered earth at the sample site, for example, hardens the spectrum in comparison with DS86 values. Even when using a fission spectrum pointing downward and neglecting air humidity one cannot explain our 36Cl measurements. If the effective thermal neutron fluences, that have a similar ratio of resonance integral to thermal neutron capture cross sections obtained from 36Cl, 41Ca and 152Eu, are averaged, a bomb yield of about 16 kt is deduced in agreement with a bomb yield of (15 +/- 3) kt estimated in DS86.

  2. ACDOS2: an improved neutron-induced dose rate code

    SciTech Connect

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  3. A measurement of the muon-induced neutron yield in lead at a depth of 2850 m water equivalent

    SciTech Connect

    Reichhart, L.; Ghag, C.; Lindote, A.; Chepel, V.; DeViveiros, L.; Lopes, M. I.; Neves, F.; Pinto da Cunha, J.; Silva, C.; Solovov, V. N.; Akimov, D. Yu.; Belov, V. A.; Burenkov, A. A.; Kobyakin, A. S.; Kovalenko, A. G.; Stekhanov, V. N.; Araújo, H. M.; Bewick, A.; Currie, A.; Horn, M.; and others

    2013-08-08

    We present results from the measurement of the neutron production rate in lead by high energy cosmic-ray muons at a depth of 2850 m water equivalent (mean muon energy of 260 GeV). A tonne-scale highly segmented plastic scintillator detector was utilised to detect both the energy depositions from the traversing muons as well as the delayed radiative capture signals of the induced neutrons. Complementary Monte Carlo simulations reproduce well the distributions of muons and detected muon-induced neutrons. Absolute agreement between simulation and data is of the order of 25%. By comparing the measured and simulated neutron capture rates a neutron yield in pure lead of (5.78{sub −0.28}{sup +0.21})×10{sup −3} neutrons/muon/(g/cm{sup 2}) has been obtained.

  4. Development and characterization of a high yield transportable pulsed neutron source with efficient and compact pulsed power system

    NASA Astrophysics Data System (ADS)

    Verma, Rishi; Mishra, Ekansh; Dhang, Prosenjit; Sagar, Karuna; Meena, Manraj; Shyam, Anurag

    2016-09-01

    The results of characterization experiments carried out on a newly developed dense plasma focus device based intense pulsed neutron source with efficient and compact pulsed power system are reported. Its high current sealed pseudospark switch based low inductance capacitor bank with maximum stored energy of ˜10 kJ is segregated into four modules of ˜2.5 kJ each and it cumulatively delivers peak current in the range of 400 kA-600 kA (corresponding to charging voltage range of 14 kV-18 kV) in a quarter time period of ˜2 μs. The neutron yield performance of this device has been optimized by discretely varying deuterium filling gas pressure in the range of 6 mbar-11 mbar at ˜17 kV/550 kA discharge. At ˜7 kJ/8.5 mbar operation, the average neutron yield has been measured to be in the order of ˜4 × 109 neutrons/pulse which is the highest ever reported neutron yield from a plasma focus device with the same stored energy. The average forward to radial anisotropy in neutron yield is found to be ˜2. The entire system is contained on a moveable trolley having dimensions 1.5 m × 1 m × 0.7 m and its operation and control (up to the distance of 25 m) are facilitated through optically isolated handheld remote console. The overall compactness of this system provides minimum proximity to small as well as large samples for irradiation. The major intended application objective of this high neutron yield dense plasma focus device development is to explore the feasibility of active neutron interrogation experiments by utilization of intense pulsed neutron sources.

  5. Neutron and gamma-ray dose-rates from the Little Boy replica

    SciTech Connect

    Plassmann, E.A.; Pederson, R.A.

    1984-01-01

    We report dose-rate information obtained at many locations in the near vicinity of, and at distances out to 0.64 km from, the Little Boy replica while it was operated as a critical assembly. The measurements were made with modified conventional dosimetry instruments that used an Anderson-Braun detector for neutrons and a Geiger-Mueller tube for gamma rays with suitable electronic modules to count particle-induced pulses. Thermoluminescent dosimetry methods provide corroborative data. Our analysis gives estimates of both neutron and gamma-ray relaxation lengths in air for comparison with earlier calculations. We also show the neutron-to-gamma-ray dose ratio as a function of distance from the replica. Current experiments and further data analysis will refine these results. 7 references, 8 figures.

  6. [Genetic changes in yeast cells Saccharomyces irradiated by fast neutrons with different dose rate].

    PubMed

    Malinova, I V; Tsyb, T S; Komarova, E V

    2009-01-01

    No neutron dose rate effects in the wide range of 10(-3) Gy/s to 10(6) Gy/s were observed in yeast diploid cells for induction of mitotic segregation and crossing-over. The RBE values for these effects were determined as doses ratio (Dgamma/D(n)) at maximum effects. The RBE were 2.2-1.9 for neutrons of the reactor BR-10 (E = = 0.85 MeV) and the pulse reactor BARS-6 (E = 1.44 MeV). The RBE values for genetic effects were 1.0 at the equal survival level for neutrons and gamma-rays 60Co.

  7. [Leakage radiations in a medical electron accelerator facility--calculation of neutron doses in the facility].

    PubMed

    Ishimatsu, K; Morikawa, K

    1990-02-01

    Neutron doses often come dominant in mazes of electron accelerator facilities in which X-rays of energies more than 10 MV are produced. A simple analytical method to calculate neutron doses in such a facility is developed. In the calculation procedure, it is assumed that the irradiation room is spherical in shape and the maze is cylindrical. Multiple reflection of neutrons is also considered using the albedo concept in the calculation. The procedure allows to exist a hanging wall over the entrance of the irradiation room and also multiple legs in the maze. All the parameters used in the calculation are given definitely in the procedure, and any experiment is unnecessary to determine value of the parameters. Comparison of the calculated results with experimental ones will be described in the following report. PMID:2326507

  8. Neutron dose per fluence and weighting factors for use at high energy accelerators

    SciTech Connect

    Cossairt, J.Donald; Vaziri, Kamran; /Fermilab

    2008-07-01

    In June 2007, the United States Department of Energy incorporated revised values of neutron weighting factors into its occupational radiation protection Regulation 10 CFR Part 835 as part of updating its radiation dosimetry system. This has led to a reassessment of neutron radiation fields at high energy proton accelerators such as those at the Fermi National Accelerator Laboratory (Fermilab). Values of dose per fluence factors appropriate for accelerator radiation fields calculated elsewhere are collated and radiation weighting factors compared. The results of this revision to the dosimetric system are applied to americium-beryllium neutron energy spectra commonly used for instrument calibrations. A set of typical accelerator neutron energy spectra previously measured at Fermilab are reassessed in light of the new dosimetry system. The implications of this revision are found to be of moderate significance.

  9. Investigation of deuterated target effects on neutron yield in plasma focus device SBUMTPF1

    NASA Astrophysics Data System (ADS)

    Shahbazi Rad, Zahra; Abbasi Davani, Fereydoun; Shirani, Babak

    2015-04-01

    In this research, the effect of inserting deuterated solid target in plasma focus device `SBUMTPF1' on neutron yield has been investigated. The deuterated target with the diameter of 2.5 cm was placed at different heights relative to the anode tip. In each height, the best place of target (where the ion density is highest) was found from observing the effects of ions struck on the aluminum samples. Also for each height, 20 shots were performed at the optimum pressure of deuterium working gas and operating voltage, which are equal to 1.5 mbar and 24 kV, respectively. The neutron production was measured with two activation counters, which placed in 0○ and 90○ relative to the anode axis. Neutron scattering from two activation counters was calculated with MCNP4C code and the results showed that this effect is negligible. In this article, the probability of implanting deuterium ions into the titanium target was also investigated. Deviation angle of the ion emission relative to the anode axis was measured experimentally in this research and it was about 3.1○.

  10. Measurement of neutron dose with an organic liquid scintillator coupled with a spectrum weight function.

    PubMed

    Kim, E; Endo, A; Yamaguchi, Y; Yoshizawa, M; Nakamura, T

    2002-01-01

    A dose evaluation method for neutrons in the energy range of a few MeV to 100 MeV has been developed using a spectrum weight function (G-function), which is applied to an organic liquid scintillator of 12.7 cm in diameter and 12.7 cm in length. The G-function that converts the pulse height spectrum of the scintillator into the ambient dose equivalent, H*(10), was calculated by an unfolding method using successive approximation of the response function of the scintillator and the ambient dose equivalent per unit neutron fluence (H*(10) conversion coefficients) of ICRP 74. To verify the response function of the scintillator and the value of H*(10) evaluated by the G-function. pulse height spectra of the scintillator were measured in some different neutron fields, which have continuous energy, monoenergetic and quasi-monoenergetic spectra. Values of H*(10) estimated using the G-function and pulse height spectra of the scintillator were compared with those calculated using neutron energy spectra. These doses agreed with each other. From the results, it was concluded that H*(10) can be evaluated directly from the pulse height spectrum of the scintillator by applying the G-function proposed in this study. PMID:12212900

  11. Low dose neutron late effects: Cataractogenesis. Final progress report, April 1, 1992--March 31, 1993

    SciTech Connect

    Worgul, B.V.

    1994-04-01

    The work is formulated to resolve the uncertainty regarding the relative biological effectiveness (RBE) of low dose neutron radiation. The study exploits the fact that cataractogenesis is sensitive to the inverse dose-rate effect as has been observed with heavy ions and was an endpoint considered in the follow-up of the A-bomb survivors. The neutron radiations were initiated at the Radiological Research Accelerator facility (RARAF) of the Nevis Laboratory of Columbia University. Four week old ({+-} 1 day) rats were divided into eight dose groups each receiving single or fractionated total doses of 0.2, 1.0, 5.0 and 25.0 cGy of monoenergetic 435 keV neutrons. Special restraining jigs insured that the eye, at the midpoint of the lens, received the appropriate energy and dose with a relative error of {+-} 5%. The fractionation regimen consisted of four exposures, each administered at three hour ({+-} 1 minute) intervals. The neutron irradiated groups were compared to rats irradiated with 250 kVp X-rays in doses ranging from 0.5 to 7 Gy. The animals were examined on a biweekly basis utilizing conventional slit-lamp biomicroscopy and the Scheimpflug Slit Lamp Imaging System (Zeiss). The follow-ups, which proceeded for over 2 years, are now complete. This proved essential inasmuch as given the extremely low doses which were utilized, clinically detectable opacities were not anticipated until a significant fraction of the life span has lapsed. The results have exceeded all expectations.

  12. Low dose neutron late effects: Cataractogenesis. Progress report, April 1, 1991--December 15, 1991

    SciTech Connect

    Worgul, B.V.

    1991-12-01

    The work is formulated to resolve the uncertainty regarding the relative biological effectiveness (RBE) of low dose neutron radiation. The study exploits the fact that cataractogenesis is sensitive to the inverse dose-rate effect as has been observed with heavy ions and was an endpoint considered in the follow-up of the A-bomb survivors. The neutron radiations were initiated at the Radiological Research Accelerator facility (RARAF) of the Nevis Laboratory of Columbia University. Four week old ({plus_minus} 1 day) rats were divided into eight dose groups each receiving single or fractionated total doses of 0.2, 1.0, 5.0 and 25.0 cGy of monoenergetic 435 KeV neutrons. Special restraining jigs insured that the eye, at the midpoint of the lens, received the appropriate energy and dose with a relative error of {plus_minus}5%. The fractionation regimen consisted of four exposures, each administered at three hour ({plus_minus}) intervals. The neutron irradiated groups are being compared to rats irradiated with 250kVp X-rays in doses ranging from 0.5 to 7 Gy. The animals are being examined on a biweekly basis utilizing conventional slit-lamp biomicroscopy and the Scheimpflug Slit Lamp Imaging System (Zeiss). The follows-ups, entering their second year, will continue throughout the life-span of the animals. This is essential inasmuch as given the extremely low doses which are being utilized clinically detectable opacities were not anticipated until a significant fraction of the life span has lapsed. Current data support this contention. At this juncture cataracts in the irradiated groups are beginning to exceed control levels.

  13. Neutron equivalent doses and associated lifetime cancer incidence risks for head & neck and spinal proton therapy

    NASA Astrophysics Data System (ADS)

    Athar, Basit S.; Paganetti, Harald

    2009-08-01

    In this work we have simulated the absorbed equivalent doses to various organs distant to the field edge assuming proton therapy treatments of brain or spine lesions. We have used computational whole-body (gender-specific and age-dependent) voxel phantoms and considered six treatment fields with varying treatment volumes and depths. The maximum neutron equivalent dose to organs near the field edge was found to be approximately 8 mSv Gy-1. We were able to clearly demonstrate that organ-specific neutron equivalent doses are age (stature) dependent. For example, assuming an 8-year-old patient, the dose to brain from the spinal fields ranged from 0.04 to 0.10 mSv Gy-1, whereas the dose to the brain assuming a 9-month-old patient ranged from 0.5 to 1.0 mSv Gy-1. Further, as the field aperture opening increases, the secondary neutron equivalent dose caused by the treatment head decreases, while the secondary neutron equivalent dose caused by the patient itself increases. To interpret the dosimetric data, we analyzed second cancer incidence risks for various organs as a function of patient age and field size based on two risk models. The results show that, for example, in an 8-year-old female patient treated with a spinal proton therapy field, breasts, lungs and rectum have the highest radiation-induced lifetime cancer incidence risks. These are estimated to be 0.71%, 1.05% and 0.60%, respectively. For an 11-year-old male patient treated with a spinal field, bronchi and rectum show the highest risks of 0.32% and 0.43%, respectively. Risks for male and female patients increase as their age at treatment time decreases.

  14. A Permanent-Magnet Microwave Ion Source for a Compact High-Yield Neutron Generator

    SciTech Connect

    Waldmann, Ole; Ludewigt, Bernhard

    2010-10-11

    We present recent work on the development of a microwave ion source that will be used in a high-yield compact neutron generator for active interrogation applications. The sealed tube generator will be capable of producing high neutron yields, 5x1011 n/s for D-T and ~;;1x1010 n/s for D-D reactions, while remaining transportable. We constructed a microwave ion source (2.45 GHz) with permanent magnets to provide the magnetic field strength of 87.5 mT necessary for satisfying the electron cyclotron resonance (ECR) condition. Microwave ion sources can produce high extracted beam currents at the low gas pressures required for sealed tube operation and at lower power levels than previously used RF-driven ion sources. A 100 mA deuterium/tritium beam will be extracted through a large slit (60x6 mm2) to spread the beam power over a larger target area. This paper describes the design of the permanent-magnet microwave ion source and discusses the impact of the magnetic field design on the source performance. The required equivalent proton beam current density of 40 mA/cm2 was extracted at a moderate microwave power of 400 W with an optimized magnetic field.

  15. Monte Carlo simulation of depth dose distribution in several organic models for boron neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Matsumoto, T.

    2007-09-01

    Monte Carlo simulations are performed to evaluate depth-dose distributions for possible treatment of cancers by boron neutron capture therapy (BNCT). The ICRU computational model of ADAM & EVA was used as a phantom to simulate tumors at a depth of 5 cm in central regions of the lungs, liver and pancreas. Tumors of the prostate and osteosarcoma were also centered at the depth of 4.5 and 2.5 cm in the phantom models. The epithermal neutron beam from a research reactor was the primary neutron source for the MCNP calculation of the depth-dose distributions in those cancer models. For brain tumor irradiations, the whole-body dose was also evaluated. The MCNP simulations suggested that a lethal dose of 50 Gy to the tumors can be achieved without reaching the tolerance dose of 25 Gy to normal tissue. The whole-body phantom calculations also showed that the BNCT could be applied for brain tumors without significant damage to whole-body organs.

  16. Measurements of neutron dose equivalent for a proton therapy center using uniform scanning proton beams

    SciTech Connect

    Zheng Yuanshui; Liu Yaxi; Zeidan, Omar; Schreuder, Andries Niek; Keole, Sameer

    2012-06-15

    Purpose: Neutron exposure is of concern in proton therapy, and varies with beam delivery technique, nozzle design, and treatment conditions. Uniform scanning is an emerging treatment technique in proton therapy, but neutron exposure for this technique has not been fully studied. The purpose of this study is to investigate the neutron dose equivalent per therapeutic dose, H/D, under various treatment conditions for uniform scanning beams employed at our proton therapy center. Methods: Using a wide energy neutron dose equivalent detector (SWENDI-II, ThermoScientific, MA), the authors measured H/D at 50 cm lateral to the isocenter as a function of proton range, modulation width, beam scanning area, collimated field size, and snout position. They also studied the influence of other factors on neutron dose equivalent, such as aperture material, the presence of a compensator, and measurement locations. They measured H/D for various treatment sites using patient-specific treatment parameters. Finally, they compared H/D values for various beam delivery techniques at various facilities under similar conditions. Results: H/D increased rapidly with proton range and modulation width, varying from about 0.2 mSv/Gy for a 5 cm range and 2 cm modulation width beam to 2.7 mSv/Gy for a 30 cm range and 30 cm modulation width beam when 18 Multiplication-Sign 18 cm{sup 2} uniform scanning beams were used. H/D increased linearly with the beam scanning area, and decreased slowly with aperture size and snout retraction. The presence of a compensator reduced the H/D slightly compared with that without a compensator present. Aperture material and compensator material also have an influence on neutron dose equivalent, but the influence is relatively small. H/D varied from about 0.5 mSv/Gy for a brain tumor treatment to about 3.5 mSv/Gy for a pelvic case. Conclusions: This study presents H/D as a function of various treatment parameters for uniform scanning proton beams. For similar treatment

  17. Graphical user interface for yield and dose estimations for cyclotron-produced technetium.

    PubMed

    Hou, X; Vuckovic, M; Buckley, K; Bénard, F; Schaffer, P; Ruth, T; Celler, A

    2014-07-01

    The cyclotron-based (100)Mo(p,2n)(99m)Tc reaction has been proposed as an alternative method for solving the shortage of (99m)Tc. With this production method, however, even if highly enriched molybdenum is used, various radioactive and stable isotopes will be produced simultaneously with (99m)Tc. In order to optimize reaction parameters and estimate potential patient doses from radiotracers labeled with cyclotron produced (99m)Tc, the yields for all reaction products must be estimated. Such calculations, however, are extremely complex and time consuming. Therefore, the objective of this study was to design a graphical user interface (GUI) that would automate these calculations, facilitate analysis of the experimental data, and predict dosimetry. The resulting GUI, named Cyclotron production Yields and Dosimetry (CYD), is based on Matlab®. It has three parts providing (a) reaction yield calculations, (b) predictions of gamma emissions and (c) dosimetry estimations. The paper presents the outline of the GUI, lists the parameters that must be provided by the user, discusses the details of calculations and provides examples of the results. Our initial experience shows that the proposed GUI allows the user to very efficiently calculate the yields of reaction products and analyze gamma spectroscopy data. However, it is expected that the main advantage of this GUI will be at the later clinical stage when entering reaction parameters will allow the user to predict production yields and estimate radiation doses to patients for each particular cyclotron run.

  18. Graphical user interface for yield and dose estimations for cyclotron-produced technetium

    NASA Astrophysics Data System (ADS)

    Hou, X.; Vuckovic, M.; Buckley, K.; Bénard, F.; Schaffer, P.; Ruth, T.; Celler, A.

    2014-07-01

    The cyclotron-based 100Mo(p,2n)99mTc reaction has been proposed as an alternative method for solving the shortage of 99mTc. With this production method, however, even if highly enriched molybdenum is used, various radioactive and stable isotopes will be produced simultaneously with 99mTc. In order to optimize reaction parameters and estimate potential patient doses from radiotracers labeled with cyclotron produced 99mTc, the yields for all reaction products must be estimated. Such calculations, however, are extremely complex and time consuming. Therefore, the objective of this study was to design a graphical user interface (GUI) that would automate these calculations, facilitate analysis of the experimental data, and predict dosimetry. The resulting GUI, named Cyclotron production Yields and Dosimetry (CYD), is based on Matlab®. It has three parts providing (a) reaction yield calculations, (b) predictions of gamma emissions and (c) dosimetry estimations. The paper presents the outline of the GUI, lists the parameters that must be provided by the user, discusses the details of calculations and provides examples of the results. Our initial experience shows that the proposed GUI allows the user to very efficiently calculate the yields of reaction products and analyze gamma spectroscopy data. However, it is expected that the main advantage of this GUI will be at the later clinical stage when entering reaction parameters will allow the user to predict production yields and estimate radiation doses to patients for each particular cyclotron run.

  19. Measurement of fission products yields in the quasi-mono-energetic neutron-induced fission of 232Th

    NASA Astrophysics Data System (ADS)

    Naik, H.; Mukherji, Sadhana; Suryanarayana, S. V.; Jagadeesan, K. C.; Thakare, S. V.; Sharma, S. C.

    2016-08-01

    The cumulative yields of various fission products in the 232Th(n, f) reaction at average neutron energies of 5.42, 7.75, 9.35 and 12.53 MeV have been determined by using an off-line γ-ray spectrometric technique. The neutron beam was produced from the 7Li(p, n) reaction by using the proton energies of 7.8, 12, 16 and 20 MeV. The mass chain yields were obtained from the cumulative fission yields by using the charge distribution correction of medium energy fission. The fine structure in the mass yield distribution was interpreted from the point of nuclear structure effect. On the other hand, the higher yield around mass number 133-134 and 143-144 as well as their complementary products were explained based on the standard I and standard II asymmetric mode of fission. From the mass yield data, the average value of light mass (), heavy mass (), the average number of neutrons (< ν >) and the peak-to-valley (P / V) ratios at different neutron energies of present work and literature data were obtained in the 232Th(n, f) reaction. The different parameters of the mass yield distribution in the 232Th(n, f) reaction were compared with the similar data in the 232Th(γ, f) reaction at comparable excitation energy and a surprising difference was observed.

  20. Determination of the cosmic-ray-induced neutron flux and ambient dose equivalent at flight altitude

    NASA Astrophysics Data System (ADS)

    Pazianotto, M. T.; Cortés-Giraldo, M. A.; Federico, C. A.; Gonçalez, O. L.; Quesada, J. M.; Carlson, B. V.

    2015-07-01

    There is interest in modeling the atmosphere in the South Atlantic Magnetic Anomaly in order to obtain information about the cosmic-ray induced neutron spectrum and angular distribution as functions of altitude. In this work we use the Monte Carlo codes MCNPX and Geant4 to determine the cosmic-ray-induced neutron flux in the atmosphere produced by the cosmic ray protons incident on the top of the atmosphere and to estimate the ambient dose equivalent rate as function of altitude. The results present a reasonable conformity to other codes (QARM and EXPACS) based on other parameterizations.

  1. Monte Carlo simulation of the neutron spectral fluence and dose equivalent for use in shielding a proton therapy vault

    PubMed Central

    Zheng, Yuanshui; Newhauser, Wayne; Klein, Eric; Low, Daniel

    2014-01-01

    Neutron production is of principal concern when designing proton therapy vault shielding. Conventionally, neutron calculations are based on analytical methods, which do not accurately consider beam shaping components and nozzle shielding. The goal of this study was to calculate, using Monte Carlo modeling, the neutron spectral fluence and neutron dose equivalent generated by a realistic proton therapy nozzle and evaluate how these data could be used in shielding calculations. We modeled a contemporary passive scattering proton therapy nozzle in detail with the MCNPX simulation code. The neutron spectral fluence and dose equivalent at various locations in the treatment room were calculated and compared to those obtained from a thick iron target bombarded by parallel proton beams, the simplified geometry on which analytical methods are based. The neutron spectral fluence distributions were similar for both methods, with deeply penetrating high-energy neutrons (E > 10 MeV) being most prevalent along the beam central axis, and low-energy neutrons predominating the neutron spectral fluence in the lateral region. However, unlike the inverse square falloff used in conventional analytical methods, this study shows that the neutron dose equivalent per therapeutic dose in the treatment room decreased with distance approximately following a power law, with an exponent of about −1.63 in the lateral region and −1.73 in the downstream region. Based on the simulated data according to the detailed nozzle modeling, we developed an empirical equation to estimate the neutron dose equivalent at any location and distance in the treatment vault, e.g. for cases in which detailed Monte Carlo modeling is not feasible. We applied the simulated neutron spectral fluence and dose equivalent to a shielding calculation as an example. PMID:19887713

  2. Monte Carlo simulation of the neutron spectral fluence and dose equivalent for use in shielding a proton therapy vault.

    PubMed

    Zheng, Yuanshui; Newhauser, Wayne; Klein, Eric; Low, Daniel

    2009-11-21

    Neutron production is of principal concern when designing proton therapy vault shielding. Conventionally, neutron calculations are based on analytical methods, which do not accurately consider beam shaping components and nozzle shielding. The goal of this study was to calculate, using Monte Carlo modeling, the neutron spectral fluence and neutron dose equivalent generated by a realistic proton therapy nozzle and evaluate how these data could be used in shielding calculations. We modeled a contemporary passive scattering proton therapy nozzle in detail with the MCNPX simulation code. The neutron spectral fluence and dose equivalent at various locations in the treatment room were calculated and compared to those obtained from a thick iron target bombarded by parallel proton beams, the simplified geometry on which analytical methods are based. The neutron spectral fluence distributions were similar for both methods, with deeply penetrating high-energy neutrons (E > 10 MeV) being most prevalent along the beam central axis, and low-energy neutrons predominating the neutron spectral fluence in the lateral region. However, unlike the inverse square falloff used in conventional analytical methods, this study shows that the neutron dose equivalent per therapeutic dose in the treatment room decreased with distance approximately following a power law, with an exponent of about -1.63 in the lateral region and -1.73 in the downstream region. Based on the simulated data according to the detailed nozzle modeling, we developed an empirical equation to estimate the neutron dose equivalent at any location and distance in the treatment vault, e.g. for cases in which detailed Monte Carlo modeling is not feasible. We applied the simulated neutron spectral fluence and dose equivalent to a shielding calculation as an example.

  3. mBAND analysis of chromosome aberrations in human epithelial cells induced by gamma-rays and secondary neutrons of low dose rate.

    PubMed

    Hada, M; Gersey, B; Saganti, P B; Wilkins, R; Cucinotta, F A; Wu, H

    2010-08-14

    Human risks from chronic exposures to both low- and high-LET radiation are of intensive research interest in recent years. In the present study, human epithelial cells were exposed in vitro to gamma-rays at a dose rate of 17 mGy/h or secondary neutrons of 25 mGy/h. The secondary neutrons have a broad energy spectrum that simulates the Earth's atmosphere at high altitude, as well as the environment inside spacecrafts like the Russian MIR station and the International Space Station (ISS). Chromosome aberrations in the exposed cells were analyzed using the multicolor banding in situ hybridization (mBAND) technique with chromosome 3 painted in 23 colored bands that allows identification of both inter- and intrachromosome exchanges including inversions. Comparison of present dose responses between gamma-rays and neutron irradiations for the fraction of cells with damaged chromosome 3 yielded a relative biological effectiveness (RBE) value of 26+/-4 for the secondary neutrons. Our results also revealed that secondary neutrons of low dose rate induced a higher fraction of intrachromosome exchanges than gamma-rays, but the fractions of inversions observed between these two radiation types were indistinguishable. Similar to the previous findings after acute radiation exposures, most of the inversions observed in the present study were accompanied by other aberrations. The fractions of complex type aberrations and of unrejoined chromosomal breakages were also found to be higher in the neutron-exposed cells than after gamma-rays. We further analyzed the location of the breaks involved in chromosome aberrations along chromosome 3, and observed hot spots after gamma-ray, but not neutron, exposures.

  4. mBAND analysis of chromosome aberrations in human epithelial cells induced by gamma-rays and secondary neutrons of low dose rate.

    PubMed

    Hada, M; Gersey, B; Saganti, P B; Wilkins, R; Cucinotta, F A; Wu, H

    2010-08-14

    Human risks from chronic exposures to both low- and high-LET radiation are of intensive research interest in recent years. In the present study, human epithelial cells were exposed in vitro to gamma-rays at a dose rate of 17 mGy/h or secondary neutrons of 25 mGy/h. The secondary neutrons have a broad energy spectrum that simulates the Earth's atmosphere at high altitude, as well as the environment inside spacecrafts like the Russian MIR station and the International Space Station (ISS). Chromosome aberrations in the exposed cells were analyzed using the multicolor banding in situ hybridization (mBAND) technique with chromosome 3 painted in 23 colored bands that allows identification of both inter- and intrachromosome exchanges including inversions. Comparison of present dose responses between gamma-rays and neutron irradiations for the fraction of cells with damaged chromosome 3 yielded a relative biological effectiveness (RBE) value of 26+/-4 for the secondary neutrons. Our results also revealed that secondary neutrons of low dose rate induced a higher fraction of intrachromosome exchanges than gamma-rays, but the fractions of inversions observed between these two radiation types were indistinguishable. Similar to the previous findings after acute radiation exposures, most of the inversions observed in the present study were accompanied by other aberrations. The fractions of complex type aberrations and of unrejoined chromosomal breakages were also found to be higher in the neutron-exposed cells than after gamma-rays. We further analyzed the location of the breaks involved in chromosome aberrations along chromosome 3, and observed hot spots after gamma-ray, but not neutron, exposures. PMID:20338263

  5. An international dosimetry exchange for boron neutron capture therapy. Part I: Absorbed dose measurements.

    PubMed

    Binns, P J; Riley, K J; Harling, O K; Kiger, W S; Munck af Rosenschöld, P M; Giusti, V; Capala, J; Sköld, K; Auterinen, I; Serén, T; Kotiluoto, P; Uusi-Simola, J; Marek, M; Viererbl, L; Spurny, F

    2005-12-01

    An international collaboration was organized to undertake a dosimetry exchange to enable the future combination of clinical data from different centers conducting neutron capture therapy trials. As a first step (Part I) the dosimetry group from the Americas, represented by MIT, visited the clinical centers at Studsvik (Sweden), VTT Espoo (Finland), and the Nuclear Research Institute (NRI) at Rez (Czech Republic). A combined VTT/NRI group reciprocated with a visit to MIT. Each participant performed a series of dosimetry measurements under equivalent irradiation conditions using methods appropriate to their clinical protocols. This entailed in-air measurements and dose versus depth measurements in a large water phantom. Thermal neutron flux as well as fast neutron and photon absorbed dose rates were measured. Satisfactory agreement in determining absorbed dose within the experimental uncertainties was obtained between the different groups although the measurement uncertainties are large, ranging between 3% and 30% depending upon the dose component and the depth of measurement. To improve the precision in the specification of absorbed dose amongst the participants, the individually measured dose components were normalized to the results from a single method. Assuming a boron concentration of 15 microg g(-1) that is typical of concentrations realized clinically with the boron delivery compound boronophenylalanine-fructose, systematic discrepancies in the specification of the total biologically weighted dose of up to 10% were apparent between the different groups. The results from these measurements will be used in future to normalize treatment plan calculations between the different clinical dosimetry protocols as Part II of this study.

  6. Effect of low doses of 14 MeV neutrons on polymers.

    PubMed

    Rivaton, Agnès; Arnold, Jack; Dos Santos, Morgane; Bussière, Pierre-Olivier; Taviot-Gueho, Christine

    2010-11-01

    The structural modifications of polymers irradiated with 14 MeV neutrons were studied. Two elastomers, a polypropylene-type polymer and poly(ethylene oxide) were exposed to low doses of fast neutrons in the range of 0.3-14 Gy. The radiation damages were observed at the molecular scale by infrared spectroscopy. The morphological changes were investigated by steric exclusion chromatography, insoluble fraction measurements, differential scanning calorimetry and X-ray diffraction. It was found that neutrons provoked oxidation processes accompanied by modifications in the polymer architecture, including chain scissions, crosslinking reactions and changes in the crystallinity. Moreover, the conventional antioxidants were shown to be inefficient in inhibiting the aging of the polymers. These results also suggest that the radiation damages could be used successfully for dosimetry applications using an easily implementable protocol.

  7. ESR investigation of joint use of dentin and tooth enamel to estimate photon and neutron dose components of a mixed field.

    PubMed

    Trompier, F; Tikunov, D D; Ivannikov, A; Clairand, I

    2006-01-01

    In the case of mixed photon and neutron field, estimation of photon and neutron dose components from the ESR signal of tooth enamel alone is impossible. To differentiate neutron and photon components using the method described in ICRU 26 for twin chambers, enamel and dentin sensitivities to photon and to neutron were investigated. Enamel and dentin relative sensitivities were, respectively, estimated at 0.03 +/- 0.02 and 0.14 +/- 0.10 for fission neutrons. Basing on this result, calculation of neutron and photon doses was performed in realistic case of criticality accident. Estimation of neutron and photon dose components was found in good agreement with reference dosimetry.

  8. Ambient neutron dose equivalent outside concrete vault rooms for 15 and 18 MV radiotherapy accelerators.

    PubMed

    Martínez-Ovalle, S A; Barquero, R; Gómez-Ros, J M; Lallena, A M

    2012-03-01

    In this work, the ambient dose equivalent, H*(10), due to neutrons outside three bunkers that house a 15- and a 18-MV Varian Clinac 2100C/D and a 15-MV Elekta Inor clinical linacs, has been calculated. The Monte Carlo code MCNPX (v. 2.5) has been used to simulate the neutron production and transport. The complete geometries including linacs and full installations have been built up according to the specifications of the manufacturers and the planes provided by the corresponding medical physical services of the hospitals where the three linacs operate. Two of these installations, those lodging the Varian linacs, have an entrance door to the bunker while the other one does not, although it has a maze with two bends. Various treatment orientations were simulated in order to establish plausible annual equivalent doses. Specifically anterior-posterior, posterior-anterior, left lateral, right lateral orientations and an additional one with the gantry rotated 30° have been studied. Significant dose rates have been found only behind the walls and the door of the bunker, near the entrance and the console, with a maximum of 12 µSv h(-1). Dose rates per year have been calculated assuming a conservative workload for the three facilities. The higher dose rates in the corresponding control areas were 799 µSv y(-1), in the case of the facility which operates the 15-MV Clinac, 159 µSv y(-1), for that with the 15-MV Elekta, and 21 µSv y(-1) for the facility housing the 18-MV Varian. A comparison with measurements performed in similar installations has been carried out and a reasonable agreement has been found. The results obtained indicate that the neutron contamination does not increase the doses above the legal limits and does not produce a significant enhancement of the dose equivalent calculated. When doses are below the detection limits provided by the measuring devices available today, MCNPX simulation provides an useful method to evaluate neutron dose equivalents based

  9. Application of the new neutron monitor yield function computed for different altitudes to an analysis of GLEs

    NASA Astrophysics Data System (ADS)

    Mishev, Alexander; Usoskin, Ilya

    2016-07-01

    A precise analysis of SEP (solar energetic particle) spectral and angular characteristics using neutron monitor (NM) data requires realistic modeling of propagation of those particles in the Earth's magnetosphere and atmosphere. On the basis of the method including a sequence of consecutive steps, namely a detailed computation of the SEP assymptotic cones of acceptance, and application of a neutron monitor yield function and convenient optimization procedure, we derived the rigidity spectra and anisotropy characteristics of several major GLEs. Here we present several major GLEs of the solar cycle 23: the Bastille day event on 14 July 2000 (GLE 59), GLE 69 on 20 January 2005, and GLE 70 on 13 December 2006. The SEP spectra and pitch angle distributions were computed in their dynamical development. For the computation we use the newly computed yield function of the standard 6NM64 neutron monitor for primary proton and alpha CR nuclei. In addition, we present new computations of NM yield function for the altitudes of 3000 m and 5000 m above the sea level The computations were carried out with Planetocosmics and CORSIKA codes as standardized Monte-Carlo tools for atmospheric cascade simulations. The flux of secondary neutrons and protons was computed using the Planetocosmics code appliyng a realistic curved atmospheric. Updated information concerning the NM registration efficiency for secondary neutrons and protons was used. The derived results for spectral and angular characteristics using the newly computed NM yield function at several altitudes are compared with the previously obtained ones using the double attenuation method.

  10. Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC.

    PubMed

    Israngkul-Na-Ayuthaya, Isra; Suriyapee, Sivalee; Pengvanich, Phongpheath

    2015-11-01

    A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed (28)Al, (24)Na, (54)Mn and (60)Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is (28)Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several (28)Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received.

  11. Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC.

    PubMed

    Israngkul-Na-Ayuthaya, Isra; Suriyapee, Sivalee; Pengvanich, Phongpheath

    2015-11-01

    A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed (28)Al, (24)Na, (54)Mn and (60)Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is (28)Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several (28)Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received. PMID:26265661

  12. Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC

    PubMed Central

    Israngkul-Na-Ayuthaya, Isra; Suriyapee, Sivalee; Pengvanich, Phongpheath

    2015-01-01

    A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed 28Al, 24Na, 54Mn and 60Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is 28Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several 28Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received. PMID:26265661

  13. Impact of switching to the ICRP-74 neutron flux-to-dose equivalent rate conversion factors at the Sandia National Laboratory Building 818 Neutron Source Range.

    SciTech Connect

    Ward, Dann C.

    2009-03-01

    Sandia National Laboratories (SNL) maintains a neutron calibration facility which supports the calibration, maintenance, and repair of Radiation Protection Instruments. The SNL neutron reference fields are calibrated using the following methodology: Fluence rate is initially established by calculation using the NIST traceable source emission rate (decay corrected). Correction factors for the effects of room return or scatter, and source anisotropy are then developed by using a suitable radiation transport code to model the geometry of the facility. The conventionally true neutron dose rates are then determined using the appropriate fluence-todose equivalent conversion coefficients at several reference positions. This report describes the impact on calculated neutron dose rates of switching from NCRP-38 to CRP-74 neutron flux-todose equivalent rate conversion factors. This switch is driven by recent changes to dosimetry requirements addressed in 10 CFR 835 (Occupational Radiation Protection).

  14. Monte Carlo modeling of proton therapy installations: a global experimental method to validate secondary neutron dose calculations.

    PubMed

    Farah, J; Martinetti, F; Sayah, R; Lacoste, V; Donadille, L; Trompier, F; Nauraye, C; De Marzi, L; Vabre, I; Delacroix, S; Hérault, J; Clairand, I

    2014-06-01

    Monte Carlo calculations are increasingly used to assess stray radiation dose to healthy organs of proton therapy patients and estimate the risk of secondary cancer. Among the secondary particles, neutrons are of primary concern due to their high relative biological effectiveness. The validation of Monte Carlo simulations for out-of-field neutron doses remains however a major challenge to the community. Therefore this work focused on developing a global experimental approach to test the reliability of the MCNPX models of two proton therapy installations operating at 75 and 178 MeV for ocular and intracranial tumor treatments, respectively. The method consists of comparing Monte Carlo calculations against experimental measurements of: (a) neutron spectrometry inside the treatment room, (b) neutron ambient dose equivalent at several points within the treatment room, (c) secondary organ-specific neutron doses inside the Rando-Alderson anthropomorphic phantom. Results have proven that Monte Carlo models correctly reproduce secondary neutrons within the two proton therapy treatment rooms. Sensitive differences between experimental measurements and simulations were nonetheless observed especially with the highest beam energy. The study demonstrated the need for improved measurement tools, especially at the high neutron energy range, and more accurate physical models and cross sections within the Monte Carlo code to correctly assess secondary neutron doses in proton therapy applications.

  15. Fast Neutron Dose Evaluation Using CR39 by Coincidence Counting Process

    SciTech Connect

    Vilela, Eudice; Freitas, F. F. de; Brandao, J. O. C.; Santos, J. A. L.

    2008-08-07

    The solid state nuclear tracks detection (SSNTD) technique is widely used in the area of radiation dosimetry. Different materials can be used applying this technique as glass and the most used in the dosimetry field that are the polycarbonates, CR39 and Makrofol-DE. Both are very rich in hydrogenous, that enables the SSNTD to detect fast neutrons through recoils of protons in the own detector material, without need of converters. The low reproducibility of its backgroundhas often been the major drawback in the assessment of low fluences of fast neutrons with SSNTDs. This problem can be effectively solved by counting coincidence of tracks in two detectors foils irradiated in close contact. After processing and counting only tracks produced by the same recoil nuclei on the surfaces of both detectors are considered as a track. This procedure enables the reduction of the background counts in the response of the detectors. In this work a preliminary study on the application of the coincidence technique for neutron dosimetry is presented. The CR39 material was investigated aiming to achieve the personal dose equivalent for fast neutrons. Using this method of analysis a significant reduction on the lower detectable dose was observed resulting even one order of magnitude smaller value. Reading, however, needs to be automated due to the large areas necessary to achieve a satisfactory number of tracks for statistical significance of results.

  16. Neutron activation analysis for reference determination of the implantation dose of cobalt ions

    SciTech Connect

    Garten, R.P.H.; Bubert, H.; Palmetshofer, L.

    1992-05-15

    The authors prepared depth profilling reference materials by cobalt ion implantation at an ion energy of 300 keV into n-type silicon. The implanted Co dose was then determined by instrumental neutron activation analysis (INAA) giving an analytical dynamic range of almost 5 decades and uncertainty of 1.5%. This form of analysis allows sources of error (beam spreading, misalignment) to be corrected. 70 refs., 3 tabs.

  17. Advantage and limitations of weighting factors and weighted dose quantities and their units in boron neutron capture therapy.

    PubMed

    Rassow, J; Sauerwein, W; Wittig, A; Bourhis-Martin, E; Hideghéty, K; Moss, R

    2004-05-01

    Defining the parameters influencing the biological reaction due to absorbed dose is a continuous topic of research. The main goal of radiobiological research is to translate the measurable dose of ionizing radiation to a quantitative expression of biological effect. Mathematical models based on different biological approaches (e.g., skin reaction, cell culture) provide some estimations that are often misleading and, to some extent, dangerous. Conventional radiotherapy is the simplest case because the primary radiation and secondary radiation are both low linear energy transfer (LET) radiation and have about the same relative biological effectiveness (RBE). Nevertheless, for this one-dose-component case, the dose-effect curves are not linear. In fact, the total absorbed dose and the absorbed dose per fraction as well as the time schedule of the fractionation scheme influence the biological effects. Mathematical models such as the linear-quadratic model can only approximate biological effects. With regard to biological effects, fast neutron therapy is more complex than conventional radiotherapy. Fast neutron beams are always contaminated by gamma rays. As a consequence, biological effects are due to two components, a high-LET component (neutrons) and a low-LET component (photons). A straight transfer of knowledge from conventional radiotherapy to fast neutron therapy is, therefore, not possible: RBE depends on the delivered dose and several other parameters. For dose reporting, the European protocol for fast neutron dosimetry recommends that the total absorbed dose with gamma-ray absorbed dose in brackets is stated. However, boron neutron capture therapy (BNCT) is an even more complex case, because the total absorbed dose is due to four dose components with different LET and RBE. In addition, the terminology and units used by the different BNCT groups is confusing: absorbed dose and weighted dose are both to be stated in grays and are never "photon equivalent." The

  18. Correlation of clinical outcome to the estimated radiation dose from Boron Neutron Capture Therapy (BNCT)

    SciTech Connect

    Chadha, M.; Coderre, J.A.; Chanana, A.D.

    1996-12-31

    A phase I/II trial delivering a single fraction of BNCT using p-Boronophenylalanine-Fructose and epithermal neutrons at the the Brookhaven Medical Research Reactor was initiated in September 1994. The primary endpiont of the study was to evaluate the feasibility and safety of a given BNCT dose. The clinical outcome of the disease was a secondary endpoint of the study. The objective of this paper is to evaluate the correlation of the clinical outcome of patients to the estimated radiation dose from BNCT.

  19. Quantitative radiation dose-response relationships for normal tissues in man - I. Gustatory tissues response during photon and neutron radiotherapy

    SciTech Connect

    Mossman, K.L.

    1982-08-01

    Quantitative radiation dose-response curves for normal gustatory tissue in man were studied. Taste function, expressed as taste loss, was evaluated in 84 patients who were given either photon or neutron radiotherapy for tumors in the head and neck region. Patients were treated to average tumor doses of 6600 cGy (photon) or 2200 cGy intervals for photon patients and 320-cGy intervals for neutron patients during radiotherapy. The dose-response curves for photons and neutrons were analyzed by fitting a four-parameter logistic equation to the data. Photon and neutron curves differed principally in their relative position along the dose axis. Comparison of the dose-response curves were made by determination of RBE. At 320 cGy, the lowest neutron dose at which taste measurements were made, RBE = 5.7. If this RBE is correct, then the therapeutic gain factor may be equal to or less than 1, indicating no biological advantage in using neutrons over photons for this normal tissue. These studies suggest measurements of taste function and evaluation of dose-response relationships may also be useful in quantitatively evaluating the efficacy of chemical modifiers of radiation response such as hypoxic cell radiosensitizers and radioprotectors.

  20. Monitoring in situ stress/strain behaviour during plastic yielding in polymineralic rocks using neutron diffraction

    NASA Astrophysics Data System (ADS)

    Covey-Crump, S. J.; Schofield, P. F.; Stretton, I. C.; Daymond, M. R.; Knight, K. S.; Tant, J.

    2013-02-01

    Attempts to use rock deformation experiments to examine the elastic and plastic behaviour of polymineralic rocks are hampered by the fact that usually only whole sample properties can be monitored as opposed to the separate contribution of each phase. To circumvent this difficulty, room-temperature, uniaxial compression experiments were performed in a neutron beam-line on a suite of calcite + halite samples with different phase volume proportions. By collecting diffraction data during loading, the elastic strain and hence stress in each phase was determined as a function of load to bulk strains of 1-2%. In all cases, the calcite behaved elastically while the halite underwent plastic yielding. During the fully elastic part of the deformation, the composite elastic properties and the within-phase stresses are well-described both by recent shear lag models and by analyses based on Eshelby's solution for the elastic field around an ellipsoidal inclusion in a homogeneous medium. After the onset of yielding, the halite in situ stress/total strain curve may be reconstructed using the rule of mixtures. At calcite contents of greater than 30%, the in situ halite response may be significantly weaker or stronger than that obtained at lesser calcite contents. The results highlight the potential that such techniques offer for developing an explicitly experimental approach for determining the influence of microstructural variables on the mechanical properties of polymineralic rocks.

  1. Determination of uranium at trace levels by radiochemical neutron-activation analysis employing radioisotopic yield evaluation.

    PubMed

    Byrne, A R; Benedik, L

    1988-03-01

    Nanogram and picogram quantities of uranium were determined in biological materials by radiochemical neutron-activation analysis. Two different approaches using either (239)U or (239)Np were employed for cross-checking, and the question of negative errors due to incomplete acid dissolution of any possible inorganic (siliceous) fraction was studied. In the first and main approach, radiochemical separation of the short-lived (239)U (23.5 min) nuclide was based on TBP extraction following rapid conventional wet-ashing. Addition of large amounts of uranium carrier (ca. 50 mg) allowed the chemical yield to be evaluated from the gamma spectrum of the isolated fraction by means of the 186 keV peak of (235)U. In the second approach, the longer-lived (239)Np (56.5 hr) daughter was separated by anion-exchange; this nuclide allowed use of lengthier dissolution procedures employing total decomposition with hydrofluoric acid. Nanogram quantities of (237)Np were irradiated simultaneously with the sample and an aliquot of the resulting solution containing (237)Np and (238)Np (51 hr) was added prior to sample destruction, these isotopes serving as carrier and yield tracer, respectively. Results are presented for a series of reference materials. The methodologies and results from the two approaches are discussed and evaluated. PMID:18964488

  2. In vivo and phantom measurements of the secondary photon and neutron doses for prostate patients undergoing 18 MV IMRT.

    PubMed

    Reft, Chester S; Runkel-Muller, Renate; Myrianthopoulos, Leon

    2006-10-01

    For intensity modulated radiation therapy (IMRT) treatments 6 MV photons are typically used, however, for deep seated tumors in the pelvic region, higher photon energies are increasingly being employed. IMRT treatments require more monitor units (MU) to deliver the same dose as conformal treatments, causing increased secondary radiation to tissues outside the treated area from leakage and scatter, as well as a possible increase in the neutron dose from photon interactions in the machine head. Here we provide in vivo patient and phantom measurements of the secondary out-of-field photon radiation and the neutron dose equivalent for 18 MV IMRT treatments. The patients were treated for prostate cancer with 18 MV IMRT at institutions using different therapy machines and treatment planning systems. Phantom exposures at the different facilities were used to compare the secondary photon and neutron dose equivalent between typical IMRT delivered treatment plans with a six field three-dimensional conformal radiotherapy (3DCRT) plan. For the in vivo measurements LiF thermoluminescent detectors (TLDs) and Al2O3 detectors using optically stimulated radiation were used to obtain the photon dose and CR-39 track etch detectors were used to obtain the neutron dose equivalent. For the phantom measurements a Bonner sphere (25.4 cm diameter) containing two types of TLDs (TLD-600 and TLD-700) having different thermal neutron sensitivities were used to obtain the out-of-field neutron dose equivalent. Our results showed that for patients treated with 18 MV IMRT the photon dose equivalent is greater than the neutron dose equivalent measured outside the treatment field and the neutron dose equivalent normalized to the prescription dose varied from 2 to 6 mSv/Gy among the therapy machines. The Bonner sphere results showed that the ratio of neutron equivalent doses for the 18 MV IMRT and 3DCRT prostate treatments scaled as the ratio of delivered MUs. We also observed differences in the

  3. Effect of long term target changes on the neutron yield from a low intensity (d, t) neutron generator

    NASA Astrophysics Data System (ADS)

    Dalton, A. W.

    1987-12-01

    Experimental and theoretical techniques have been developed to determine the accuracy with which the integrated neutron output from a low-intensity (d, t) neutron source can be measured during a prolonged irradiation. The experiments involved a neutron generator in which a fixed solid titanium-tritium target and an unanalysed beam of deuterium ions was used. The analysis was based on differential and integral measurements of both the deuterium beam current and the energy spectra of the charged particles emitted from the multiple nuclear interactions in the target during beam bombardment. The overlapping signals produced by the latter are interpreted using an iterative analysis developed at the Lucas Heights Laboratories.

  4. Monte Carlo neutron doses estimations inside a PET cyclotron vault room.

    PubMed

    Barquero, R; Méndez, R; Martí-Climent, J M; Quincoces, G

    2007-01-01

    Neutron organ equivalent doses, effective doses and dose equivalents received inside a positron emission tomography vault room in a maximum credible accident have been estimated with the Monte Carlo code MCNPX. While an operator was inside the vault room of a Cyclone 18/9 IBA cyclotron, this was producing (18)F with 30 muA proton current in the target and the operator had to activate a stopped emergency device placed on the wall. MC simulation of the cyclotron vault were carried out to estimate the organ and tissue equivalent doses in a mathematical male mannequin simulating the operator facing the wall on which the emergency device is placed. Doses were calculated at two emergency devices for each one of the two targets of the cyclotron, which were able to produce (18)F. The maximum effective dose in the mannequin was 6.70 Sv/h and the maximum organ equivalent dose was 18.47 Sv/h in spleen.

  5. Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

    NASA Astrophysics Data System (ADS)

    Sharma, Manish K.; Alajo, Ayodeji B.

    2016-08-01

    Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

  6. Comparison of out-of-field photon doses in 6 MV IMRT and neutron doses in proton therapy for adult and pediatric patients

    NASA Astrophysics Data System (ADS)

    Athar, Basit S.; Bednarz, Bryan; Seco, Joao; Hancox, Cindy; Paganetti, Harald

    2010-05-01

    The purpose of this study was to assess lateral out-of-field doses in 6 MV IMRT (intensity modulated radiation therapy) and compare them with secondary neutron equivalent dose contributions in proton therapy. We simulated out-of-field photon doses to various organs as a function of distance, patient's age, gender and treatment volumes based on 3, 6, 9 cm field diameters in the head and neck and spine region. The out-of-field photon doses to organs near the field edge were found to be in the range of 2, 5 and 10 mSv Gy-1 for 3 cm, 6 cm and 9 cm diameter IMRT fields, respectively, within 5 cm of the field edge. Statistical uncertainties calculated in organ doses vary from 0.2% to 40% depending on the organ location and the organ volume. Next, a comparison was made with previously calculated neutron equivalent doses from proton therapy using identical field arrangements. For example, out-of-field doses for IMRT to lung and uterus (organs close to the 3 cm diameter spinal field) were computed to be 0.63 and 0.62 mSv Gy-1, respectively. These numbers are found to be a factor of 2 smaller than the corresponding out-of-field doses for proton therapy, which were estimated to be 1.6 and 1.7 mSv Gy-1 (RBE), respectively. However, as the distance to the field edge increases beyond approximately 25 cm the neutron equivalent dose from proton therapy was found to be a factor of 2-3 smaller than the out-of-field photon dose from IMRT. We have also analyzed the neutron equivalent doses from an ideal scanned proton therapy (assuming not significant amount of absorbers in the treatment head). Out-of-field doses were found to be an order of magnitude smaller compared to out-of-field doses in IMRT or passive scattered proton therapy. In conclusion, there seem to be three geometrical areas when comparing the out-of-target dose from IMRT and (passive scattered) proton treatments. Close to the target (in-field, not analyzed here) protons offer a distinct advantage due to the lower

  7. Comparison of out-of-field photon doses in 6 MV IMRT and neutron doses in proton therapy for adult and pediatric patients

    PubMed Central

    Athar, Basit S; Bednarz, Bryan; Seco, Joao; Hancox, Cindy; Paganetti, Harald

    2012-01-01

    The purpose of this study was to assess lateral out-of-field doses in 6 MV IMRT (intensity modulated radiation therapy) and compare them with secondary neutron equivalent dose contributions in proton therapy. We simulated outof-field photon doses to various organs as a function of distance, patient's age, gender and treatment volumes based on 3, 6, 9 cm field diameters in the head and neck and spine region. The out-of-field photon doses to organs near the field edge were found to be in the range of 2, 5 and 10 mSv Gy−1 for 3 cm, 6 cm and 9 cm diameter IMRT fields, respectively, within 5 cm of the field edge. Statistical uncertainties calculated in organ doses vary from 0.2% to 40% depending on the organ location and the organ volume. Next, a comparison was made with previously calculated neutron equivalent doses from proton therapy using identical field arrangements. For example, out-of-field doses for IMRT to lung and uterus (organs close to the 3 cm diameter spinal field) were computed to be 0.63 and 0.62 mSv Gy−1, respectively. These numbers are found to be a factor of 2 smaller than the corresponding out-of-field doses for proton therapy, which were estimated to be 1.6 and 1.7 mSv Gy−1 (RBE), respectively. However, as the distance to the field edge increases beyond approximately 25 cm the neutron equivalent dose from proton therapy was found to be a factor of 2–3 smaller than the out-of-field photon dose from IMRT. We have also analyzed the neutron equivalent doses from an ideal scanned proton therapy (assuming not significant amount of absorbers in the treatment head). Outof-field doses were found to be an order of magnitude smaller compared to out-of-field doses in IMRT or passive scattered proton therapy. In conclusion, there seem to be three geometrical areas when comparing the out-of-target dose from IMRT and (passive scattered) proton treatments. Close to the target (in-field, not analyzed here) protons offer a distinct advantage due to the lower

  8. The effect of neutron irradiation dose on vacancy defect accumulation and annealing in pure nickel

    NASA Astrophysics Data System (ADS)

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

    2012-02-01

    In order to investigate the dose dependence of vacancy defect evolution in nickel, specimens of high-purity Ni 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. Ni was 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 concentration of 3D-VCs increases with increasing dose level. 3D-VCs collapse into secondary-type clusters (stacking fault tetrahedra (SFTs), and vacancy loops) during stepwise annealing at 350-450 K. It is shown that the thermal stability of SFTs grow with increasing dose level, probably, it is due to growth of the average SFT size during annealing. The results of annealing experiments on electron-irradiated Ni at 300 K are indicated in the paper, for comparison. We also have briefly discussed the positron response to the SFT-like structures.

  9. Measurement of neutron dose equivalent outside and inside of the treatment vault of GRID therapy

    SciTech Connect

    Wang, Xudong; Charlton, Michael A.; Esquivel, Carlos; Eng, Tony Y.; Li, Ying; Papanikolaou, Nikos

    2013-09-15

    Purpose: To evaluate the neutron and photon dose equivalent rates at the treatment vault entrance (H{sub n,D} and H{sub G}), and to study the secondary radiation to the patient in GRID therapy. The radiation activation on the grid was studied.Methods: A Varian Clinac 23EX accelerator was working at 18 MV mode with a grid manufactured by .decimal, Inc. The H{sub n,D} and H{sub G} were measured using an Andersson–Braun neutron REM meter, and a Geiger Müller counter. The radiation activation on the grid was measured after the irradiation with an ion chamber γ-ray survey meter. The secondary radiation dose equivalent to patient was evaluated by etched track detectors and OSL detectors on a RANDO{sup ®} phantom.Results: Within the measurement uncertainty, there is no significant difference between the H{sub n,D} and H{sub G} with and without a grid. However, the neutron dose equivalent to the patient with the grid is, on average, 35.3% lower than that without the grid when using the same field size and the same amount of monitor unit. The photon dose equivalent to the patient with the grid is, on average, 44.9% lower. The measured average half-life of the radiation activation in the grid is 12.0 (±0.9) min. The activation can be categorized into a fast decay component and a slow decay component with half-lives of 3.4 (±1.6) min and 15.3 (±4.0) min, respectively. There was no detectable radioactive contamination found on the surface of the grid through a wipe test.Conclusions: This work indicates that there is no significant change of the H{sub n,D} and H{sub G} in GRID therapy, compared with a conventional external beam therapy. However, the neutron and scattered photon dose equivalent to the patient decrease dramatically with the grid and can be clinical irrelevant. Meanwhile, the users of a grid should be aware of the possible high dose to the radiation worker from the radiation activation on the surface of the grid. A delay in handling the grid after the beam

  10. Effect of diameter of nanoparticles and capture cross-section library on macroscopic dose enhancement in boron neutron capture therapy

    PubMed Central

    Farhood, Bagher

    2014-01-01

    Purpose The aim of this study is evaluation of the effect of diameter of 10B nanoparticles and various neutron capture cross-section libraries on macroscopic dose enhancement in boron neutron capture therapy (BNCT). Material and methods MCNPX Monte Carlo code was used for simulation of a 252Cf source, a soft tissue phantom and a tumor containing 10B nanoparticles. Using 252Cf as a neutron source, macroscopic dose enhancement factor (MDEF) and total dose rate in tumor in the presence of 100, 200, and 500 ppm of 10B nanoparticles with 25 nm, 50 nm, and 100 nm diameters were calculated. Additionally, the effect of ENDF, JEFF, JENDL, and CENDL neutron capture cross-section libraries on MDEF was evaluated. Results There is not a linear relationship between the average MDEF value and nanoparticles’ diameter but the average MDEF grows with increased concentration of 10B nanoparticles. There is an increasing trend for average MDEF with the tumor distance. The average MDEF values were obtained the same for various neutron capture cross-section libraries. The maximum and minimum doses that effect on the total dose in tumor were neutron and secondary photon doses, respectively. Furthermore, the boron capture related dose component reduced in some extent with increase of diameter of 10B nanoparticles. Conclusions Based on the results of this study, it can be concluded that from physical point of view, various nanoparticle diameters have no dominant effect on average MDEF value in tumor. Furthermore, it is concluded that various neutron capture cross-section libraries are resulted to the same macroscopic dose enhancements. However, it is predicted that taking into account the biological effects for various nanoparticle diameters will result in different dose enhancements. PMID:25834582

  11. Determination of neutron dose from criticality accidents with bioassays for sodium-24 in blood and phosphorus-32 in hair

    SciTech Connect

    Feng, Y.; Miller, L.F.; Brown, K.S.; Casson, W.H.; Mei, G.T.; Thein, M.

    1993-06-01

    A comprehensive review of accident neutron dosimetry using blood and hair analysis was performed and is summarized in this report. Experiments and calculations were conducted at Oak Ridge National Laboratory (ORNL) and the University of Tennessee (UT) to develop measurement techniques for the activity of {sup 24}Na in blood and {sup 32}P in hair for nuclear accident dosimetry. An operating procedure was established for the measurement of {sup 24}Na in blood using an HPGe detector system. The sensitivity of the measurement for a 20-mL sample is 0.01-0.02 Gy of total neutron dose for hard spectra and below 0.005 Gy for soft spectra based on a 30- to 60-min counting time. The operating procedures for direct counting of hair samples are established using a liquid scintillation detector. Approximately 0.06-0.1 Gy of total neutron dose can be measured from a 1-g hair sample using this procedure. Detailed procedures for chemical dissolution and ashing of hair samples are also developed. A method is proposed to use blood and hair analysis for assessing neutron dose based on a collection of 98 neutron spectra. Ninety-eight blood activity-to-dose conversion factors were calculated. The calculated results for an uncollided fission spectrum compare favorably with previously published data for fission neutrons. This nuclear accident dosimetry system makes it possible to estimate an individual`s neutron dose within a few hours after an accident if the accident spectrum can be approximated from one of 98 tabulated neutron spectrum descriptions. If the information on accident and spectrum description is not available, the activity ratio of {sup 32}P in hair and {sup 24}Na in blood can provide information related to the neutron spectrum for dose assessment.

  12. Feasibility study of the neutron dose for real-time image-guided proton therapy: A Monte Carlo study

    NASA Astrophysics Data System (ADS)

    Kim, Jin Sung; Shin, Jung Suk; Kim, Daehyun; Shin, Eunhyuk; Chung, Kwangzoo; Cho, Sungkoo; Ahn, Sung Hwan; Ju, Sanggyu; Chung, Yoonsun; Jung, Sang Hoon; Han, Youngyih

    2015-07-01

    Two full rotating gantries with different nozzles (multipurpose nozzle with MLC, scanning dedicated nozzle) for a conventional cyclotron system are installed and being commissioned for various proton treatment options at Samsung Medical Center in Korea. The purpose of this study is to use Monte Carlo simulation to investigate the neutron dose equivalent per therapeutic dose, H/D, for X-ray imaging equipment under various treatment conditions. At first, we investigated the H/D for various modifications of the beamline devices (scattering, scanning, multi-leaf collimator, aperture, compensator) at the isocenter and at 20, 40 and 60 cm distances from the isocenter, and we compared our results with those of other research groups. Next, we investigated the neutron dose at the X-ray equipment used for real-time imaging under various treatment conditions. Our investigation showed doses of 0.07 ~ 0.19 mSv/Gy at the X-ray imaging equipment, depending on the treatment option and interestingly, the 50% neutron dose reduction was observed due to multileaf collimator during proton scanning treatment with the multipurpose nozzle. In future studies, we plan to measure the neutron dose experimentally and to validate the simulation data for X-ray imaging equipment for use as an additional neutron dose reduction method.

  13. Monte Carlo Simulations on Neutron Transport and Absorbed Dose in Tissue-Equivalent Phantoms Exposed to High-Flux Epithermal Neutron Beams

    NASA Astrophysics Data System (ADS)

    Bartesaghi, G.; Gambarini, G.; Negri, A.; Carrara, M.; Burian, J.; Viererbl, L.

    2010-04-01

    Presently there are no standard protocols for dosimetry in neutron beams for boron neutron capture therapy (BNCT) treatments. Because of the high radiation intensity and of the presence at the same time of radiation components having different linear energy transfer and therefore different biological weighting factors, treatment planning in epithermal neutron fields for BNCT is usually performed by means of Monte Carlo calculations; experimental measurements are required in order to characterize the neutron source and to validate the treatment planning. In this work Monte Carlo simulations in two kinds of tissue-equivalent phantoms are described. The neutron transport has been studied, together with the distribution of the boron dose; simulation results are compared with data taken with Fricke gel dosimeters in form of layers, showing a good agreement.

  14. Testing JEFF-3.1.1 and ENDF/B-VII.1 Decay and Fission Yield Nuclear Data Libraries with Fission Pulse Neutron Emission and Decay Heat Experiments

    NASA Astrophysics Data System (ADS)

    Cabellos, O.; de Fusco, V.; Diez de la Obra, C. J.; Martinez, J. S.; Gonzalez, E.; Cano-Ott, D.; Alvarez-Velarde, F.

    2014-04-01

    The aim of this work is to test the present status of Evaluated Nuclear Decay and Fission Yield Data Libraries to predict decay heat and delayed neutron emission rate, average neutron energy and neutron delayed spectra after a neutron fission pulse. Calculations are performed with JEFF-3.1.1 and ENDF/B-VII.1, and these are compared with experimental values. An uncertainty propagation assessment of the current nuclear data uncertainties is performed.

  15. Monte-Carlo calculations of particle fluences and neutron effective dose rates in the atmosphere.

    PubMed

    Matthiä, Daniel; Sihver, Lembit; Meier, Matthias

    2008-01-01

    Monitoring of radiation exposure of aircrew is a legal requirement for many airlines in the EU and a challenging task in dosimetry. Monte-Carlo simulations of cosmic particles in the atmosphere can contribute to the understanding of the corresponding radiation field. Calculations of secondary neutron fluences in the atmosphere produced by galactic cosmic rays together with the resulting neutron-effective dose rates are shown in this paper and compared with results from the AIR project. The PLANETOCOSMICS package based on GEANT4 and two models for the local interstellar spectra of galactic cosmic rays have been used for the calculations. Furthermore, secondary muon fluences have been computed and are compared with CAPRICE measurements.

  16. Neutron and gamma-ray dose measurements at various distances from the Little Boy replica

    SciTech Connect

    Huntzinger, C.J.; Hankins, D.E.

    1984-08-01

    We measured neutron and gamma-ray dose rates at various distances from the Little Boy-Comet Critical Assembly at Los Alamos National Laboratory (LANL) in April of 1983. The Little Boy-Comet Assembly is a replica of the atomic weapon detonated over Hiroshima, designed to be operated at various steady-state power levels. The selected distances for measurement ranged from 107 m to 567 m. Gamma-ray measurements were made with a Reuter-Stokes environmental ionization chamber which has a sensitivity of 1.0 ..mu..R/hour. Neutron measurements were made with a pulsed-source remmeter which has a sensitivity of 0.1 ..mu..rem/hour, designed and built at Lawrence Livermore National Laboratory (LLNL). 12 references, 7 figures, 6 tables.

  17. Neutron dose rate at the SwissFEL injector test facility: first measurements.

    PubMed

    Hohmann, E; Frey, N; Fuchs, A; Harm, C; Hödlmoser, H; Lüscher, R; Mayer, S; Morath, O; Philipp, R; Rehmann, A; Schietinger, T

    2014-10-01

    At the Paul Scherrer Institute, the new SwissFEL Free Electron Laser facility is currently in the design phase. It is foreseen to accelerate electrons up to a maximum energy of 7 GeV with a pulsed time structure. An injector test facility is operated at a maximum energy of 300 MeV and serves as the principal test and demonstration plant for the SwissFEL project. Secondary radiation is created in unavoidable interactions of the primary beam with beamline components. The resulting ambient dose-equivalent rate due to neutrons was measured along the beamline with different commercially available survey instruments. The present study compares the readings of these neutron detectors (one of them is specifically designed for measurements in pulsed fields). The experiments were carried out in both, a normal and a diagnostic mode of operation of the injector.

  18. Effects of the neutral gas density distribution in a DPF neutron yield

    SciTech Connect

    Milanese, M.; Moroso, R.; Pouzo, J.

    1996-12-31

    The dense plasma gives an average neutron yield Y = 2 {times} 10{sup 8} when it is operated using D{sub 2} at an homogeneous pressure p = 1.5 mb in the discharge chamber, in this p-static operation, the frequency of good shots (Y > 10{sup 7}) is f {approx} 50%. In this work the authors show the results on Y and f when PACO is operated in gas-puff way with two different modalities: (1) A gas cloud is injected into the vacuum, from a set of holes distributed in a diameter of the inner electrode near the Pyrex insulator. The gas is introduced from the back of the hollow inner electrode by means of a fast valve. The cloud expands in the interelectrode space, and reaches the extreme of the coaxial cavity in some hundreds of microseconds from the valve aperture instant. In this way of operation the value of Y remains similar to the p-static operation, but the frequency f is improved up to f {approx} 80%. (2) With a relative low value of p in the discharge chamber a jet of high density D{sub 2} is injected along the axis from the inner electrode just in the focus zone. The jet is produced with a nozzle designed in order to obtain subsonic velocity, and the gas is injected through the same fast valve. In this jet operation mode the PACO performance was improved, reaching f {approx} 70% and Y {approx} 10{sup 9}.

  19. SU-E-T-567: Neutron Dose Equivalent Evaluation for Pencil Beam Scanning Proton Therapy with Apertures

    SciTech Connect

    Geng, C; Schuemann, J; Moteabbed, M; Paganetti, H

    2015-06-15

    Purpose: To determine the neutron contamination from the aperture in pencil beam scanning during proton therapy. Methods: A Monte Carlo based proton therapy research platform TOPAS and the UF-series hybrid pediatric phantoms were used to perform this study. First, pencil beam scanning (PBS) treatment pediatric plans with average spot size of 10 mm at iso-center were created and optimized for three patients with and without apertures. Then, the plans were imported into TOPAS. A scripting method was developed to automatically replace the patient CT with a whole body phantom positioned according to the original plan iso-center. The neutron dose equivalent was calculated using organ specific quality factors for two phantoms resembling a 4- and 14-years old patient. Results: The neutron dose equivalent generated by the apertures in PBS is 4–10% of the total neutron dose equivalent for organs near the target, while roughly 40% for organs far from the target. Compared to the neutron dose equivalent caused by PBS without aperture, the results show that the neutron dose equivalent with aperture is reduced in the organs near the target, and moderately increased for those organs located further from the target. This is due to the reduction of the proton dose around the edge of the CTV, which causes fewer neutrons generated in the patient. Conclusion: Clinically, for pediatric patients, one might consider adding an aperture to get a more conformal treatment plan if the spot size is too large. This work shows the somewhat surprising fact that adding an aperture for beam scanning for facilities with large spot sizes reduces instead of increases a potential neutron background in regions near target. Changran Geng is supported by the Chinese Scholarship Council (CSC) and the National Natural Science Foundation of China (Grant No. 11475087)

  20. Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations.

    PubMed

    Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

    2014-01-01

    High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.

  1. Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations

    PubMed Central

    Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

    2014-01-01

    High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by 241Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy-1 for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy-1 achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production. PMID:24600167

  2. Dose-Dependent Onset of Regenerative Program in Neutron Irradiated Mouse Skin

    PubMed Central

    Artibani, Mara; Kobos, Katarzyna; Colautti, Paolo; Negri, Rodolfo; Amendola, Roberto

    2011-01-01

    Background Tissue response to irradiation is not easily recapitulated by cell culture studies. The objective of this investigation was to characterize, the transcriptional response and the onset of regenerative processes in mouse skin irradiated with different doses of fast neutrons. Methodology/Principal Findings To monitor general response to irradiation and individual animal to animal variation, we performed gene and protein expression analysis with both pooled and individual mouse samples. A high-throughput gene expression analysis, by DNA oligonucleotide microarray was done with three months old C57Bl/6 mice irradiated with 0.2 and 1 Gy of mono-energetic 14 MeV neutron compared to sham irradiated controls. The results on 440 irradiation modulated genes, partially validated by quantitative real time RT-PCR, showed a dose-dependent up-regulation of a sub-class of keratin and keratin associated proteins, and members of the S100 family of Ca2+-binding proteins. Immunohistochemistry confirmed mRNA expression data enabled mapping of protein expression. Interestingly, proteins up-regulated in thickening epidermis: keratin 6 and S100A8 showed the most significant up-regulation and the least mouse-to-mouse variation following 0.2 Gy irradiation, in a concerted effort toward skin tissue regeneration. Conversely, mice irradiated at 1 Gy showed most evidence of apoptosis (Caspase-3 and TUNEL staining) and most 8-oxo-G accumulation at 24 h post-irradiation. Moreover, no cell proliferation accompanied 1 Gy exposure as shown by Ki67 immunohistochemistry. Conclusions/Significance The dose-dependent differential gene expression at the tissue level following in vivo exposure to neutron radiation is reminiscent of the onset of re-epithelialization and wound healing and depends on the proportion of cells carrying multiple chromosomal lesions in the entire tissue. Thus, this study presents in vivo evidence of a skin regenerative program exerted independently from DNA repair

  3. The effect of a paraffin screen on the neutron dose at the maze door of a 15 MV linear accelerator

    SciTech Connect

    Krmar, M.; Kuzmanović, A.; Nikolić, D.; Kuzmanović, Z.; Ganezer, K.

    2013-08-15

    Purpose: The purpose of this study was to explore the effects of a paraffin screen located at various positions in the maze on the neutron dose equivalent at the maze door.Methods: The neutron dose equivalent was measured at the maze door of a room containing a 15 MV linear accelerator for x-ray therapy. Measurements were performed for several positions of the paraffin screen covering only 27.5% of the cross-sectional area of the maze. The neutron dose equivalent was also measured at all screen positions. Two simple models of the neutron source were considered in which the first assumed that the source was the cross-sectional area at the inner entrance of the maze, radiating neutrons in an isotropic manner. In the second model the reduction in the neutron dose equivalent at the maze door due to the paraffin screen was considered to be a function of the mean values of the neutron fluence and energy at the screen.Results: The results of this study indicate that the equivalent dose at the maze door was reduced by a factor of 3 through the use of a paraffin screen that was placed inside the maze. It was also determined that the contributions to the dosage from areas that were not covered by the paraffin screen as viewed from the dosimeter, were 2.5 times higher than the contributions from the covered areas. This study also concluded that the contributions of the maze walls, ceiling, and floor to the total neutron dose equivalent were an order of magnitude lower than those from the surface at the far end of the maze.Conclusions: This study demonstrated that a paraffin screen could be used to reduce the neutron dose equivalent at the maze door by a factor of 3. This paper also found that the reduction of the neutron dose equivalent was a linear function of the area covered by the maze screen and that the decrease in the dose at the maze door could be modeled as an exponential function of the product φ·E at the screen.

  4. Computation of Radiation Dose at Aircraft Altitudes from Analysis of Cosmic Ray Neutron Monitor Data

    NASA Astrophysics Data System (ADS)

    Smart, D. F.; Shea, M. A.

    Relativistic solar proton events GLEs those events with protons having sufficient kinetic energy to initiate a nuclear cascade in the atmosphere can make a contribution to radiation dose at aircraft altitudes We show that it is possible to obtain proper estimates of the expected radiation dose at aircraft altitudes from the analysis of ground-level neutron monitor data Assuming a nominal GLE spectrum the radiation dose at 40 000 feet during a 100 increase at polar latitudes will be in the range of 5 to 10 micro Sieverts per hour depending on the spectral slope An analysis of the large GLE s that have occurred during the past two solar cycles shows that there have been no events where the hourly averaged radiation dose at 40 000 feet would have exceeded 20 micro Sieverts per hour In the past improper GLE analysis was used to estimate the radiation dose at aircraft altitudes The old values derived for the early GLE s resulted in the prediction of high dose rates that persist today as urban legends and contribute to the public concept that the radiation dose at aircraft altitudes is dangerous We demonstrate that modern analytical techniques result in computed radiation doses during high-energy solar cosmic ray events that are orders of magnitude lower than those obtained by the old techniques We show that the use of the old technique of using straight line power law spectra to extrapolate the flux derived at 1 GeV results in order of magnitude errors when these flux values are extrapolated to lower energies and used to

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

    NASA Astrophysics Data System (ADS)

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

    2011-11-01

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

  6. Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

    SciTech Connect

    Finn, Erin C.; Metz, Lori A.; Greenwood, Lawrence R.; Pierson, Bruce; Wittman, Richard S.; Friese, Judah I.; Kephart, Rosara F.

    2015-04-09

    The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

  7. SU-E-T-568: Neutron Dose Survey of a Compact Single Room Proton Machine

    SciTech Connect

    Chen, Y; Prusator, M; Islam, M; Johnson, D; Ahmad, S

    2015-06-15

    Purpose: To ensure acceptable radiation limits are maintained for those working at and near the machine during its operation, a comprehensive radiation survey was performed prior to the clinical release of Mevion S250 compact proton machine at Stephenson Oklahoma Cancer Center. Methods: The Mevion S250 proton therapy system consists of the following: a superconducting cyclotron to accelerate the proton particles, a passive double scattering system for beam shaping, and paired orthogonal x-ray imaging systems for patient setup and verification via a 6D robotic couch. All equipment is housed within a single vault of compact design. Two beam delivery applicators are available for patient treatment, offering field sizes of as great as 14 cm and 25 cm in diameter, respectively. Typical clinical dose rates are between 1 and 2 Gy/min with a fixed beam energy of 250 MeV. The large applicator (25 cm in diameter) was used in conjunction with a custom cut brass aperture to create a 20 cm x 20 cm field size at beam isocenter. A 30 cm − 30 cm − 35 cm high density plastic phantom was placed in the beam path to mimic the conditions creating patient scatter. Measurements integrated-ambient-neutron-dose-equivalence were made with a SWENDII detector. Gantry angles of 0, 90 and 180 degrees, with a maximum dose rate of 150 MU/min (for large applicator) and beam configuration of option 1 (range 25 cm and 20 cm modulation), were selected as testing conditions. At each point of interest, the highest reading was recorded at 30 cm from the barrier surface. Results: The highest neutron dose was estimated to be 0.085 mSv/year at the console area. Conclusion: All controlled areas are under 5 mSv/year and the uncontrolled areas are under 1 mSv/year. The radiation protection provided by the proton vault is of sufficient quality.

  8. Annular shape silver lined proportional counter for on-line pulsed neutron yield measurement

    NASA Astrophysics Data System (ADS)

    Dighe, P. M.; Das, D.

    2015-04-01

    An annular shape silver lined proportional counter is developed to measure pulsed neutron radiation. The detector has 314 mm overall length and 235 mm overall diameter. The central cavity of 150 mm diameter and 200 mm length is used for placing the neutron source. Because of annular shape the detector covers >3π solid angle of the source. The detector has all welded construction. The detector is developed in two halves for easy mounting and demounting. Each half is an independent detector. Both the halves together give single neutron pulse calibration constant of 4.5×104 neutrons/shot count. The detector operates in proportional mode which gives enhanced working conditions in terms of dead time and operating range compared to Geiger Muller based neutron detectors.

  9. Measurement of the muon-induced neutron yield in liquid scintillator and stainless steel at LNGS with the LVD experiment

    SciTech Connect

    Persiani, R.; Garbini, M.; Sartorelli, G.; Selvi, M.; Collaboration: LVD Collaboration

    2013-08-08

    We describe the measurement of the muon-induced neutron yield in liquid scintillator and stainless steel (SS) at the Gran Sasso National Laboratory (LNGS), with the LVD experiment. The Large Volume Detector (LVD) is located in Hall A of the LNGS and is made of 1000 t of liquid scintillator and 1000 t of SS. Using an independent measurement to evaluate the background and with the support of a full Monte Carlo simulation based on Geant4, we measured a neutron yield of (2.9±0.6)×10{sup −4} and (1.5±0.3)×10{sup −3} in liquid scintillator and in stainless steel, respectively.

  10. Assessment of neutron fluence to organ dose conversion coefficients in the ORNL analytical adult phantom.

    PubMed

    Miri Hakimabad, H; Rafat Motavalli, L; Karimi Shahri, K

    2009-03-01

    Neutron fluence to absorbed dose conversion coefficients have been evaluated for the analytical ORNL modified adult phantom in 21 body organs using MCNP4C Monte Carlo code. The calculation used 20 monodirectional monoenergetic neutron beams in the energy range 10(-9)-20 MeV, under four irradiation conditions: anterior-posterior (AP), posterior-anterior (PA), left-lateral (LLAT) and right-lateral (RLAT). Then the conversion coefficients are compared with the data reported in ICRP publication 74 for mathematical MIRD type phantoms and by Bozkurt et al for the VIPMAN voxel model. Although the ORNL results show fewer differences with the ICRP results than the Bozkurt et al data, one can deduce neither complete agreement nor disparity between this study and other data sets. This comparison shows that in some cases any differences in applied Monte Carlo codes or simulated body models could significantly change the organ dose conversion factors. This sensitivity should be considered for radiological protection programmes. For certain organs, the results of two models with major differences can be in a satisfactory agreement because of the similarity in those organ models. PMID:19225185

  11. Photo neutron dose equivalent rate in 15 MV X-ray beam from a Siemens Primus Linac

    PubMed Central

    Ghasemi, A.; Pourfallah, T. Allahverdi; Akbari, M. R.; Babapour, H.; Shahidi, M.

    2015-01-01

    Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER) utilizing conversion factors of American Association of Physicist in Medicine's (AAPM) report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10-6 (3 m from isocenter in +Y direction, 0 × 0 field size) and 8.36 × 10-8 Sv/min (in maze, 40 × 40 field size), respectively. For fast neutrons, maximum DERs using 9” and 3” moderators were 1.6 × 10-5 and 1.74 × 10-5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size), respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons. PMID:26170555

  12. Assessment of organ-specific neutron equivalent doses in proton therapy using computational whole-body age-dependent voxel phantoms

    NASA Astrophysics Data System (ADS)

    Zacharatou Jarlskog, Christina; Lee, Choonik; Bolch, Wesley E.; Xu, X. George; Paganetti, Harald

    2008-02-01

    Proton beams used for radiotherapy will produce neutrons when interacting with matter. The purpose of this study was to quantify the equivalent dose to tissue due to secondary neutrons in pediatric and adult patients treated by proton therapy for brain lesions. Assessment of the equivalent dose to organs away from the target requires whole-body geometrical information. Furthermore, because the patient geometry depends on age at exposure, age-dependent representations are also needed. We implemented age-dependent phantoms into our proton Monte Carlo dose calculation environment. We considered eight typical radiation fields, two of which had been previously used to treat pediatric patients. The other six fields were additionally considered to allow a systematic study of equivalent doses as a function of field parameters. For all phantoms and all fields, we simulated organ-specific equivalent neutron doses and analyzed for each organ (1) the equivalent dose due to neutrons as a function of distance to the target; (2) the equivalent dose due to neutrons as a function of patient age; (3) the equivalent dose due to neutrons as a function of field parameters; and (4) the ratio of contributions to secondary dose from the treatment head versus the contribution from the patient's body tissues. This work reports organ-specific equivalent neutron doses for up to 48 organs in a patient. We demonstrate quantitatively how organ equivalent doses for adult and pediatric patients vary as a function of patient's age, organ and field parameters. Neutron doses increase with increasing range and modulation width but decrease with field size (as defined by the aperture). We analyzed the ratio of neutron dose contributions from the patient and from the treatment head, and found that neutron-equivalent doses fall off rapidly as a function of distance from the target, in agreement with experimental data. It appears that for the fields used in this study, the neutron dose lateral to the

  13. Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

    SciTech Connect

    Ku, L.; Kolibal, J.G.

    1982-06-01

    The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented.

  14. Neutron Energy Spectra and Yields from the 7Li(p,n) Reaction for Nuclear Astrophysics

    NASA Astrophysics Data System (ADS)

    Tessler, M.; Friedman, M.; Schmidt, S.; Shor, A.; Berkovits, D.; Cohen, D.; Feinberg, G.; Fiebiger, S.; Krása, A.; Paul, M.; Plag, R.; Plompen, A.; Reifarth, R.

    2016-01-01

    Neutrons produced by the 7Li(p, n)7Be reaction close to threshold are widely used to measure the cross section of s-process nucleosynthesis reactions. While experiments have been performed so far with Van de Graaff accelerators, the use of RF accelerators with higher intensities is planned to enable investigations on radioactive isotopes. In parallel, high-power Li targets for the production of high-intensity neutrons at stellar energies are developed at Goethe University (Frankfurt, Germany) and SARAF (Soreq NRC, Israel). However, such setups pose severe challenges for the measurement of the proton beam intensity or the neutron fluence. In order to develop appropriate methods, we studied in detail the neutron energy distribution and intensity produced by the thick-target 7Li(p,n)7Be reaction and compared them to state-of- the-art simulation codes. Measurements were performed with the bunched and chopped proton beam at the Van de Graaff facility of the Institute for Reference Materials and Measurements (IRMM) using the time-of-flight (TOF) technique with thin (1/8") and thick (1") detectors. The importance of detailed simulations of the detector structure and geometry for the conversion of TOF to a neutron energy is stressed. The measured neutron spectra are consistent with those previously reported and agree well with Monte Carlo simulations that include experimentally determined 7Li(p,n) cross sections, two-body kinematics and proton energy loss in the Li-target.

  15. Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

    SciTech Connect

    Masashi Shimada; M. Hara; T. Otsuka; Y. Oya; Y. Hatano

    2014-05-01

    Accurately estimating tritium retention in plasma facing components (PFCs) and minimizing its uncertainty are key safety issues for licensing future fusion power reactors. D-T fusion reactions produce 14.1 MeV neutrons that activate PFCs and create radiation defects throughout the bulk of the material of these components. Recent studies show that tritium migrates and is trapped in bulk (>> 10 µm) tungsten beyond the detection range of nuclear reaction analysis technique [1-2], and thermal desorption spectroscopy (TDS) technique becomes the only established diagnostic that can reveal hydrogen isotope behavior in in bulk (>> 10 µm) tungsten. Radiation damage and its recovery mechanisms in neutron-irradiated tungsten are still poorly understood, and neutron-irradiation data of tungsten is very limited. In this paper, systematic investigations with repeated plasma exposures and thermal desorption are performed to study defect annealing and thermal desorption of deuterium in low dose neutron-irradiated tungsten. Three tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to high flux (ion flux of (0.5-1.0)x1022 m-2s-1 and ion fluence of 1x1026 m-2) deuterium plasma at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy (TDS) was performed with a ramp rate of 10 °C/min up to 900 °C, and the samples were annealed at 900 °C for 0.5 hour. These procedures were repeated three (for 100 and 200 °C samples) and four (for 500 °C sample) times to uncover damage recovery mechanisms and its effects on deuterium behavior. The results show that deuterium retention decreases approximately 90, 75, and 66 % for 100, 200, and 500 °C, respectively after each annealing. When subjected to the same TDS recipe, the desorption temperature shifts from 800 °C to 600 °C after 1st annealing

  16. Cancer radiotherapy based on femtosecond IR laser-beam filamentation yielding ultra-high dose rates and zero entrance dose.

    PubMed

    Meesat, Ridthee; Belmouaddine, Hakim; Allard, Jean-François; Tanguay-Renaud, Catherine; Lemay, Rosalie; Brastaviceanu, Tiberius; Tremblay, Luc; Paquette, Benoit; Wagner, J Richard; Jay-Gerin, Jean-Paul; Lepage, Martin; Huels, Michael A; Houde, Daniel

    2012-09-18

    Since the invention of cancer radiotherapy, its primary goal has been to maximize lethal radiation doses to the tumor volume while keeping the dose to surrounding healthy tissues at zero. Sadly, conventional radiation sources (γ or X rays, electrons) used for decades, including multiple or modulated beams, inevitably deposit the majority of their dose in front or behind the tumor, thus damaging healthy tissue and causing secondary cancers years after treatment. Even the most recent pioneering advances in costly proton or carbon ion therapies can not completely avoid dose buildup in front of the tumor volume. Here we show that this ultimate goal of radiotherapy is yet within our reach: Using intense ultra-short infrared laser pulses we can now deposit a very large energy dose at unprecedented microscopic dose rates (up to 10(11) Gy/s) deep inside an adjustable, well-controlled macroscopic volume, without any dose deposit in front or behind the target volume. Our infrared laser pulses produce high density avalanches of low energy electrons via laser filamentation, a phenomenon that results in a spatial energy density and temporal dose rate that both exceed by orders of magnitude any values previously reported even for the most intense clinical radiotherapy systems. Moreover, we show that (i) the type of final damage and its mechanisms in aqueous media, at the molecular and biomolecular level, is comparable to that of conventional ionizing radiation, and (ii) at the tumor tissue level in an animal cancer model, the laser irradiation method shows clear therapeutic benefits. PMID:22927378

  17. Personal dose equivalent conversion coefficients for neutron fluence over the energy range of 20 to 250 MeV

    SciTech Connect

    Mclean, Thomas D; Justus, Alan L; Gadd, S Milan; Olsher, Richard H; Devine, Robert T

    2009-01-01

    Monte Carlo simulations were performed to extend existing neutron personal dose equivalent fluence-to-dose conversion coefficients to an energy of 250 MeV. Presently, conversion coefficients, H(p,slab)(10,alpha)/Phi, are given by ICRP-74 and ICRU-57 for a range of angles of radiation incidence (alpha = 0, 15, 30, 45, 60 and 75 degrees ) in the energy range from thermal to 20 MeV. Standard practice has been to base operational dose quantity calculations <20 MeV on the kerma approximation, which assumes that charged particle secondaries are locally deposited, or at least that charged particle equilibrium exists within the tally cell volume. However, with increasing neutron energy the kerma approximation may no longer be valid for some energetic secondaries such as protons. The Los Alamos Monte Carlo radiation transport code MCNPX was used for all absorbed dose calculations. Transport models and collision-based energy deposition tallies were used for neutron energies >20 MeV. Both light and heavy ions (HIs) (carbon, nitrogen and oxygen recoil nuclei) were transported down to a lower energy limit (1 keV for light ions and 5 MeV for HIs). Track energy below the limit was assumed to be locally deposited. For neutron tracks <20 MeV, kerma factors were used to obtain absorbed dose. Results are presented for a discrete set of angles of incidence on an ICRU tissue slab phantom.

  18. Personal dose equivalent conversion coefficients for neutron fluence over the energy range of 20-250 MeV.

    PubMed

    Olsher, R H; McLean, T D; Justus, A L; Devine, R T; Gadd, M S

    2010-03-01

    Monte Carlo simulations were performed to extend existing neutron personal dose equivalent fluence-to-dose conversion coefficients to an energy of 250 MeV. Presently, conversion coefficients, H(p,slab)(10,alpha)/Phi, are given by ICRP-74 and ICRU-57 for a range of angles of radiation incidence (alpha = 0, 15, 30, 45, 60 and 75 degrees ) in the energy range from thermal to 20 MeV. Standard practice has been to base operational dose quantity calculations <20 MeV on the kerma approximation, which assumes that charged particle secondaries are locally deposited, or at least that charged particle equilibrium exists within the tally cell volume. However, with increasing neutron energy the kerma approximation may no longer be valid for some energetic secondaries such as protons. The Los Alamos Monte Carlo radiation transport code MCNPX was used for all absorbed dose calculations. Transport models and collision-based energy deposition tallies were used for neutron energies >20 MeV. Both light and heavy ions (HIs) (carbon, nitrogen and oxygen recoil nuclei) were transported down to a lower energy limit (1 keV for light ions and 5 MeV for HIs). Track energy below the limit was assumed to be locally deposited. For neutron tracks <20 MeV, kerma factors were used to obtain absorbed dose. Results are presented for a discrete set of angles of incidence on an ICRU tissue slab phantom. PMID:19887515

  19. Progress in obtaining an absolute calibration of a total deuterium-tritium neutron yield diagnostic based on copper activationa)

    NASA Astrophysics Data System (ADS)

    Ruiz, C. L.; Chandler, G. A.; Cooper, G. W.; Fehl, D. L.; Hahn, K. D.; Leeper, R. J.; McWatters, B. R.; Nelson, A. J.; Smelser, R. M.; Snow, C. S.; Torres, J. A.

    2012-10-01

    The 350-keV Cockroft-Walton accelerator at Sandia National laboratory's Ion Beam facility is being used to calibrate absolutely a total DT neutron yield diagnostic based on the 63Cu(n,2n)62Cu(β+) reaction. These investigations have led to first-order uncertainties approaching 5% or better. The experiments employ the associated-particle technique. Deuterons at 175 keV impinge a 2.6 μm thick erbium tritide target producing 14.1 MeV neutrons from the T(d,n)4He reaction. The alpha particles emitted are measured at two angles relative to the beam direction and used to infer the neutron flux on a copper sample. The induced 62Cu activity is then measured and related to the neutron flux. This method is known as the F-factor technique. Description of the associated-particle method, copper sample geometries employed, and the present estimates of the uncertainties to the F-factor obtained are given.

  20. Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of (14)C by LSC.

    PubMed

    Bartoníček, B; Kučera, J; Světlík, I; Viererbl, L; Lahodová, Z; Tomášková, L; Cabalka, M

    2014-11-01

    Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the (14)N(n,p) (14)C reaction on nitrogen present in alanine and subsequent measurement of (14)C by liquid scintillation counting (LSC).

  1. Effect of irradiation of wheat grains with fast neutrons on the grain yield and other characteristics of the plants.

    PubMed

    Hanafy, Magda S; Mohamed, Hanan A

    2014-04-01

    The effects of fast neutrons from a (252)Cf source in the fluence range 10(5)-10(8)n/cm(2) on the Egyptian wheat cultivar (Sakha 92) were studied. The experiment was conducted for three successive seasons (2008/2009, 2009/2010, and 2010/2011) to study the effect of the irradiation on the plant growth, grain yield, and physiological changes of three generations of plants produced by irradiated moisturized grains. A low fast-neutron fluence 2 × 10(6)n/cm(2) increased the yield throughout the three mutagenic generations considerably. It also increased concentrations of the total chlorophyll, sugars, and crude protein. These changes improve the quantity and quality of the grain. Also, a study of the effect of salinity of the irrigation water on the characteristics of the third-generation grains produced by neutron-irradiated grains was performed. With increasing concentration of sodium chloride in the irrigation water in the range 0.5-1.5%, concentrations of osmoprotectants, namely, reducing sugars and proline amino acids, increased. The concentration of Na(+) in the grains increased in parallel with the salinity of the irrigation water regardless of irradiation, while the concentrations of Ca(2+) and K(+) decreased. PMID:24509363

  2. Effect of irradiation of wheat grains with fast neutrons on the grain yield and other characteristics of the plants.

    PubMed

    Hanafy, Magda S; Mohamed, Hanan A

    2014-04-01

    The effects of fast neutrons from a (252)Cf source in the fluence range 10(5)-10(8)n/cm(2) on the Egyptian wheat cultivar (Sakha 92) were studied. The experiment was conducted for three successive seasons (2008/2009, 2009/2010, and 2010/2011) to study the effect of the irradiation on the plant growth, grain yield, and physiological changes of three generations of plants produced by irradiated moisturized grains. A low fast-neutron fluence 2 × 10(6)n/cm(2) increased the yield throughout the three mutagenic generations considerably. It also increased concentrations of the total chlorophyll, sugars, and crude protein. These changes improve the quantity and quality of the grain. Also, a study of the effect of salinity of the irrigation water on the characteristics of the third-generation grains produced by neutron-irradiated grains was performed. With increasing concentration of sodium chloride in the irrigation water in the range 0.5-1.5%, concentrations of osmoprotectants, namely, reducing sugars and proline amino acids, increased. The concentration of Na(+) in the grains increased in parallel with the salinity of the irrigation water regardless of irradiation, while the concentrations of Ca(2+) and K(+) decreased.

  3. Neutron yields for reactions induced by 120 GeV protons on thick copper target

    SciTech Connect

    Kajimoto, Tsuyoshi; Sanami, Toshiya; Iwamoto, Yosuke; Shigyo, Nobuhiro; Hagiwara, Masayuki; Saitoh, Kiwamu; Nakashima, Hiroshi; Ishibashi, Kenji; Lee, Hee-Seock; Ramberg, Eric; Coleman, Richard; /Fermilab

    2011-02-01

    We developed an experimental method to measure neutron energy spectrum for 120-GeV protons on a thick copper target at Fermilab Test Beam Facility (FTBF). The spectrum in the energy range from 16 to 1600 MeV was obtained for 60-cm long copper target by time-of-flight technique with an NE213 scintillator and 5.5-m flight path. Energy spectra of neutrons generated from an interaction with beam and materials are important to design shielding structure of high energy accelerators. Until now, the energy spectra for the incident energy up to 3 GeV have been measured by several groups, Ishibashi et al., Amian et al., and Leray et al. In the energy region above 3 GeV, few experimental data are available because of small number of facilities for neutron experiment. On the other hand, concerning simulation codes, theoretical models for particle generation and transportation are switched from intermediate to high energy one around this energy. The spectra calculated by the codes have not been examined using experimental data. In shielding experiments using 120 GeV hadron beam, experimental data shows systematic differences from calculations. Hagiwara et al. have measured leakage neutron spectra behind iron and concrete shield from 120 GeV proton on target at anti-proton target station in Fermilab by using Bonner Spheres with unfolding technique. In CERN, Nakao et al reported experimental results of neutron spectra behind iron and concrete wall from 120 GeV/c proton and pion mixed beam on copper by using NE213 liquid scintillators with unfolding technique. Both of the results reported systematic discrepancies between experimental and calculation results. Therefore, experimental data are highly required to verify neutron production part of calculations. In this study, we developed an experimental method to measure neutron energy spectrum for 120 GeV proton on target. The neutron energy was determined using time-of-flight technique. We used the Fermilab Test Beam Facility (FTBF

  4. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    SciTech Connect

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Wirth, Brian D; Snead, Lance Lewis

    2016-01-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

  5. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Snead, Lance L.; Wirth, Brian D.

    2016-03-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (∼90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S-W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage, providing insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

  6. Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

    NASA Astrophysics Data System (ADS)

    McArthur, Matthew S.; Rees, Lawrence B.; Czirr, J. Bart

    2016-08-01

    Using the combination of a neutron-sensitive 6Li glass scintillator detector with a neutron-insensitive 7Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on 6Li. We used this detector with a 252Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

  7. Boron self-shielding effects on dose delivery of neutron capture therapy using epithermal beam and boronophenylalanine.

    PubMed

    Ye, S J

    1999-11-01

    Previous dosimetry studies for boron neutron capture therapy have often neglected the thermal neutron self-shielding effects caused by the 10B accumulation in the brain and the tumor. The neglect of thermal neutron flux depression, therefore, results in an overestimation of the actual dose delivery. The relevant errors are expected to be more pronounced when boronophenylalanine is used in conjunction with an epithermal neutron beam. In this paper, the boron self-shielding effects are calculated in terms of the thermal neutron flux depression across the brain and the dose delivered to the tumors. The degree of boron self-shielding is indicated by the difference between the thermal neutron fluxes calculated with and without considering a 10B concentration as part of the head phantom composition. The boron self-shielding effect is found to increase with increasing 10B concentrations and penetration depths from the skin. The calculated differences for 10B concentrations of 7.5-30 ppm are 2.3%-8.3% at 2.3 cm depth (depth of the maximum brain dose) and 4.6%-17% at 7.3 cm depth (the center of the brain). The additional self-shielding effects by the 10B concentration in a bulky tumor are investigated for a 3-cm-diam spherical tumor located either near the surface (3.3 cm depth) or at the center of the brain (7.3 cm depth) along the beam centerline. For 45 ppm of 10B in the tumor and 15 ppm of 10B in the brain, the dose delivered to the tumors is approximately 10% lower at 3.3 cm depth and 20% lower at the center of the brain, compared to the dose neglecting the boron self-shielding in transport calculations.

  8. Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method.

    2000-11-16

    Version 03 This package includes the SKYNEUT 1.1, SKYDOSE 2.3, MCSKY 2.3 and SKYCONES 1.1 codes plus the DLC-188/SKYDATA library to form a comprehensive system for calculating skyshine doses. See the author's web site for related information: http://athena.mne.ksu.edu/~jks/ SKYNEUT evaluates the neutron and neutron-induced secondary gamma-ray skyshine doses from an isotropic, point, neutron source collimated by three simple geometries: an open silo, a vertical black (perfectly absorbing) wall, and a rectangular building. The source maymore » emit monoenergetic neutrons or neutrons with an arbitrary multigroup spectrum of energies. SKYDOSE evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated by three simple geometries: (1) a source in a silo, (2) a source behind an infinitely long, vertical, black wall, and (3) a source in a rectangular building. In all three geometries an optional overhead slab shield may be specified. MCSKY evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated into either a vertical cone (i.e., silo geometry) or into a vertically oriented structure with an N-sided polygon cross section. An overhead laminate shield composed of two different materials is assumed, although shield thicknesses of zero may be specified to model an unshielded SKYSHINE source. SKYCONES evaluates the skyshine doses produced by a point neutron or gamma-photon source emitting, into the atmosphere, radiation that is collimated into an upward conical annulus between two arbitrary polar angles. The source is assumed to be axially (azimuthally) symmetric about a vertical axis through the source and can have an arbitrary polyenergetic spectrum. Nested contiguous annular cones can thus be used to represent the energy and polar-angle dependence of a skyshine source emitting radiation into the atmosphere.« less

  9. Measurements of absolute delayed neutron yield and group constants in the fast fission of {sup 235}U and {sup 237}Np

    SciTech Connect

    Loaiza, D.J.; Brunson, G.; Sanchez, R.; Butterfield, K.

    1998-03-01

    The delayed neutron activity resulting from the fast induced fission of {sup 235}U and {sup 237}Np has been studied. The six-group decay constants, relative abundances, and absolute yield of delayed neutrons from fast fission of {sup 235}U and {sup 237}Np were measured using the Godiva IV fast assembly at the Los Alamos Critical Experiments Facility. The absolute yield measured for {sup 235}U was 0.0163 {+-} 0.0008 neutron/fission. This value compares very well with the well-established Keepin absolute yield of 0.0165 {+-} 0.0005. The absolute yield value measured for {sup 237}Np was 0.0126 {+-} 0.0007. The measured delayed neutron parameters for {sup 235}U are corroborated with period (e-folding time) versus reactivity calculations.

  10. Dose conversion coefficients for neutron exposure to the lens of the human eye

    SciTech Connect

    Manger, Ryan P; Bellamy, Michael B; Eckerman, Keith F

    2011-01-01

    Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10{sup -9} to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

  11. Neutron yield study of direct-drive, low-adiabat cryogenic D{sub 2} implosions on OMEGA laser system

    SciTech Connect

    Hu, S. X.; Radha, P. B.; Marozas, J. A.; Betti, R.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Goncharov, V. N.; Igumenshchev, I. V.; Marshall, F. J.; McCrory, R. L.; Meyerhofer, D. D.; Regan, S. P.; Sangster, T. C.; Skupsky, S.; Smalyuk, V. A.; Elbaz, Y.; Shvarts, D.

    2009-11-15

    Neutron yields of direct-drive, low-adiabat ({alpha}{approx_equal}2 to 3) cryogenic D{sub 2} target implosions on the OMEGA laser system [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)] have been systematically investigated using the two-dimensional (2D) radiation hydrodynamics code DRACO[P. B. Radha et al., Phys. Plasmas 12, 056307 (2005)]. Low-mode (l{<=}12) perturbations, including initial target offset, ice-layer roughness, and laser-beam power imbalance, were found to be the primary source of yield reduction for thin-shell (5 {mu}m), low-{alpha}, cryogenic targets. The 2D simulations of thin-shell implosions track experimental measurements for different target conditions and peak laser intensities ranging from 2.5x10{sup 14}-6x10{sup 14} W/cm{sup 2}. Simulations indicate that the fusion yield is sensitive to the relative phases between the target offset and the ice-layer perturbations. The results provide a reasonable good guide to understanding the yield degradation in direct-drive, low-adiabat, cryogenic, thin-shell-target implosions. Thick-shell (10 {mu}m) implosions generally give lower yield over clean than low-l-mode DRACO simulation predictions. Simulations including the effect of laser-beam nonuniformities indicate that high-l-mode perturbations caused by laser imprinting further degrade the neutron yield of thick-shell implosions. To study ICF compression physics, these results suggest a target specification with a {<=}30 {mu}m offset and ice-roughness of {sigma}{sub rms}<3 {mu}m are required.

  12. Method for measuring dose-equivalent in a neutron flux with an unknown energy spectra and means for carrying out that method

    DOEpatents

    Distenfeld, Carl H.

    1978-01-01

    A method for measuring the dose-equivalent for exposure to an unknown and/or time varing neutron flux which comprises simultaneously exposing a plurality of neutron detecting elements of different types to a neutron flux and combining the measured responses of the various detecting elements by means of a function, whose value is an approximate measure of the dose-equivalent, which is substantially independent of the energy spectra of the flux. Also, a personnel neutron dosimeter, which is useful in carrying out the above method, comprising a plurality of various neutron detecting elements in a single housing suitable for personnel to wear while working in a radiation area.

  13. The Response of a Mouse Sarcoma to Single and Divided Doses of X-rays and Fast Neutrons

    PubMed Central

    Denekamp, J.

    1974-01-01

    The response of an experimental sarcoma to single doses and two fractions of x-rays and fast neutrons has been investigated to test the hypothesis that slowly shrinking sarcomata will reoxygenate poorly and therefore will benefit more from fractionated neutron treatment than from fractionated x-ray treatment, in contrast with rapidly shrinking carcinomata. Neutrons were approximately three times more effective than x-rays, both for single doses and for two fractions given in 48 hours, when regrowth was used as a measure of response. This observation is closely similar to results previously obtained on a rat fibrosarcoma and contrasts with previous results from a mouse mammary carcinoma, and is in agreement with the hypothesis. PMID:4855057

  14. Optimum design of a moderator system based on dose calculation for an accelerator driven Boron Neutron Capture Therapy.

    PubMed

    Inoue, R; Hiraga, F; Kiyanagi, Y

    2014-06-01

    An accelerator based BNCT has been desired because of its therapeutic convenience. However, optimal design of a neutron moderator system is still one of the issues. Therefore, detailed studies on materials consisting of the moderator system are necessary to obtain the optimal condition. In this study, the epithermal neutron flux and the RBE dose have been calculated as the indicators to look for optimal materials for the filter and the moderator. As a result, it was found that a combination of MgF2 moderator with Fe filter gave best performance, and the moderator system gave a dose ratio greater than 3 and an epithermal neutron flux over 1.0×10(9)cm(-2)s(-1).

  15. Evaluation of the spectrometric and dose characteristics of neutron fields inside the Russian segment of the ISS by fission detectors

    NASA Astrophysics Data System (ADS)

    Shurshakov, V. A.; Vorob'ev, I. B.; Nikolaev, V. A.; Lyagushin, V. I.; Akatov, Yu. A.; Kushin, V. V.

    2016-03-01

    The results of measuring the dose and the energy spectrum of neutrons inside the Russian segment of the International Space Station (ISS) from March 21 until November 10, 2002 are presented. Statistically reliable results of measurement are obtained by using thorium- and uranium-based fission detectors with cadmium and boron filters. The kits of the detectors with filters have been arranged in three compartments within assembled passive detectors in the BRADOS space experiment. The ambient dose rate H* = 139 μSv day and an energy spectrum of neutrons in the range of 10-2-104 MeV is obtained as average for the ISS compartments and is compared with the measurements carried out inside the compartments of the MIR space station. Recommendations on how to improve the procedure for using the fission detectors to measure the characteristics of neutron fields inside the compartments of space stations are formulated.

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

    PubMed

    Walsh, Linda

    2013-03-01

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

  17. Characterization of Neutron and Gamma Dose in the Irradiation Cell of Texas A and M University Research Reactor

    SciTech Connect

    Vasudevan, Latha; Reece, Warren D.; Chirayath, Sunil S.; Aghara, Sukesh

    2011-07-01

    The Monte Carlo N-Particle (MCNP) code was used to develop a three dimensional computational model of the Texas A and M University Nuclear Science Center Reactor (NSCR) operating against the irradiation (dry cell) at steady state thermal power of 1 MW. The geometry of the NSCR core and the dry cell were modeled in detail. NSCR is used for a wide variety of experiments that utilizes the dry cell for neutron as well as gamma irradiation of samples. Information on the neutron and gamma radiation environment inside the dry cell is required to facilitate irradiation of samples. This paper presents the computed neutron flux, neutron and gamma dose rate, and foil reaction rates in the dry cell, obtained through MCNP5 simulations of the NSCR core. The neutron flux was measured using foil activation method and the reaction rates obtained from {sup 197}Au(n,{gamma}){sup 198}Au and {sup 54}Fe(n,p){sup 54}Mn were compared with the model and they showed agreement within {approx} 20%. The gamma dose rate at selected locations inside the dry cell was measured using radiochromic films and the results indicate slightly higher dose rates than predicted from the model. This is because the model calculated only prompt gamma dose rates during reactor operation while the radiochromic films measured gammas from activation products and fission product decayed gammas. The model was also used to calculate the neutron energy spectra for the energy range from 0.001 eV- 20 MeV. (authors)

  18. Off-axis dose equivalent due to secondary neutrons from uniform scanning proton beams during proton radiotherapy.

    PubMed

    Islam, M R; Collums, T L; Zheng, Y; Monson, J; Benton, E R

    2013-11-21

    The production of secondary neutrons is an undesirable byproduct of proton therapy and it is important to quantify the contribution from secondary neutrons to patient dose received outside the treatment volume. The purpose of this study is to investigate the off-axis dose equivalent from secondary neutrons experimentally using CR-39 plastic nuclear track detectors (PNTD) at ProCure Proton Therapy Center, Oklahoma City, OK. In this experiment, we placed several layers of CR-39 PNTD laterally outside the treatment volume inside a phantom and in air at various depths and angles with respect to the primary beam axis. Three different proton beams with max energies of 78, 162 and 226 MeV and 4 cm modulation width, a 5 cm diameter brass aperture, and a small snout located 38 cm from isocenter were used for the entire experiment. Monte Carlo simulations were also performed based on the experimental setup using a simplified snout configuration and the FLUKA Monte Carlo radiation transport code. The measured ratio of secondary neutron dose equivalent to therapeutic primary proton dose (H/D) ranged from 0.3 ± 0.08 mSv Gy−1 for 78 MeV proton beam to 37.4 ± 2.42 mSv Gy−1 for 226 MeV proton beam. Both experiment and simulation showed a similar decreasing trend in dose equivalent with distance to the central axis and the magnitude varied by a factor of about 2 in most locations. H/D was found to increase as the energy of the primary proton beam increased and higher H/D was observed at 135° compared to 45° and 90°. The overall higher H/D in air indicates the predominance of external neutrons produced in the nozzle rather than inside the body.

  19. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    SciTech Connect

    Franco, Manuel,

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

  20. SU-E-T-108: Development of a Novel Clinical Neutron Dose Monitor for Proton Therapy Based On Twin TLD500 Chips in a Small PE Moderator

    SciTech Connect

    Hentschel, R; Mukherjee, B

    2014-06-01

    Purpose: In proton therapy, it could be desirable to measure out-of-field fast neutron doses at critical locations near and outside the patient body. Methods: The working principle of a novel clinical neutron dose monitor is verified by MCNPX simulation. The device is based on a small PE moderator of just 5.5cm side length for easy handling covered with a thermal neutron suppression layer. In the simulation, a polystyrene phantom is bombarded with a standard proton beam. The secondary thermal neutron flux produced inside the moderator by the impinging fast neutrons from the treatment volume is estimated by pairs of α-Al2O3:C (TLD500) chips which are evaluated offline after the treatment either by TL or OSL methods. The first chip is wrapped with 0.5mm natural Gadolinium foil converting the thermal neutrons to gammas via (n,γ) reaction. The second chip is wrapped with a dummy material. The chip centers have a distance of 2cm from each other. Results: The simulation shows that the difference of gamma doses in the TLD500 chips is correlated to the mean fast neutron dose delivered to the moderator material. Different outer shielding materials have been studied. 0.5mm Cadmium shielding is preferred for cost reasons and convenience. Replacement of PE moderator material by other materials like lead or iron at any place is unfavorable. The spatial orientation of the moderator cube is uncritical. Using variance reduction techniques like splitting/Russian roulette, the TLD500 gamma dose simulation give positive differences up to distances of 0.5m from the treatment volume. Conclusion: Applicability and basic layout of a novel clinical neutron dose monitor are demonstrated. The monitor measures PE neutron doses at locations outside the patient body up to distances of 0.5m from the treatment volume. Tissue neutron doses may be calculated using neutron kerma factors.

  1. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    DOE PAGES

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Wirth, Brian D; Snead, Lance Lewis

    2016-01-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutronmore » irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.« less

  2. Macroscopic geometric heterogeneity effects in radiation dose distribution analysis for boron neutron capture therapy

    SciTech Connect

    Moran, J.M.; Nigg, D.W.; Wheeler, F.J.; Bauer, W.F. )

    1992-05-01

    Calculations of radiation flux and dose distributions for boron neutron capture therapy (BNCT) of brain tumors are typically performed using sophisticated three-dimensional analytical models based on either a homogeneous approximation or a simplified few-region approximation to the actual highly heterogeneous geometry of the irradiation volume. Such models should be validated by comparison with calculations using detailed models in which all significant macroscopic tissue heterogeneities and geometric structures are explicitly represented as faithfully as possible. This paper describes such a validation exercise for BNCT of canine brain tumors. Geometric measurements of the canine anatomical structures of interest for this work were performed by dissecting and examining two essentially identical Labrador retriever heads. Chemical analyses of various tissue samples taken during the dissections were conducted to obtain measurements of elemental compositions for the tissues of interest. The resulting geometry and tissue composition data were then used to construct a detailed heterogeneous calculational model of the Labrador head. Calculations of three-dimensional radiation flux distributions pertinent to BNCT were performed for this model using the TORT discrete-ordinates radiation transport code. The calculations were repeated for a corresponding volume-weighted homogeneous-tissue model. Comparison of the results showed that peak neutron and photon flux magnitudes were quite similar for the two models (within 5%), but that the spatial flux profiles were shifted in the heterogeneous model such that the fluxes in some locations away from the peak differed from the corresponding fluxes in the homogeneous model by as much as 10%--20%. Differences of this magnitude can be therapeutically significant, emphasizing the need for proper validation of simplified treatment planning models.

  3. PHITS simulations of absorbed dose out-of-field and neutron energy spectra for ELEKTA SL25 medical linear accelerator

    NASA Astrophysics Data System (ADS)

    Puchalska, Monika; Sihver, Lembit

    2015-06-01

    Monte Carlo (MC) based calculation methods for modeling photon and particle transport, have several potential applications in radiotherapy. An essential requirement for successful radiation therapy is that the discrepancies between dose distributions calculated at the treatment planning stage and those delivered to the patient are minimized. It is also essential to minimize the dose to radiosensitive and critical organs. With MC technique, the dose distributions from both the primary and scattered photons can be calculated. The out-of-field radiation doses are of particular concern when high energy photons are used, since then neutrons are produced both in the accelerator head and inside the patients. Using MC technique, the created photons and particles can be followed and the transport and energy deposition in all the tissues of the patient can be estimated. This is of great importance during pediatric treatments when minimizing the risk for normal healthy tissue, e.g. secondary cancer. The purpose of this work was to evaluate 3D general purpose PHITS MC code efficiency as an alternative approach for photon beam specification. In this study, we developed a model of an ELEKTA SL25 accelerator and used the transport code PHITS for calculating the total absorbed dose and the neutron energy spectra infield and outside the treatment field. This model was validated against measurements performed with bubble detector spectrometers and Boner sphere for 18 MV linacs, including both photons and neutrons. The average absolute difference between the calculated and measured absorbed dose for the out-of-field region was around 11%. Taking into account a simplification for simulated geometry, which does not include any potential scattering materials around, the obtained result is very satisfactorily. A good agreement between the simulated and measured neutron energy spectra was observed while comparing to data found in the literature.

  4. PHITS simulations of absorbed dose out-of-field and neutron energy spectra for ELEKTA SL25 medical linear accelerator.

    PubMed

    Puchalska, Monika; Sihver, Lembit

    2015-06-21

    Monte Carlo (MC) based calculation methods for modeling photon and particle transport, have several potential applications in radiotherapy. An essential requirement for successful radiation therapy is that the discrepancies between dose distributions calculated at the treatment planning stage and those delivered to the patient are minimized. It is also essential to minimize the dose to radiosensitive and critical organs. With MC technique, the dose distributions from both the primary and scattered photons can be calculated. The out-of-field radiation doses are of particular concern when high energy photons are used, since then neutrons are produced both in the accelerator head and inside the patients. Using MC technique, the created photons and particles can be followed and the transport and energy deposition in all the tissues of the patient can be estimated. This is of great importance during pediatric treatments when minimizing the risk for normal healthy tissue, e.g. secondary cancer. The purpose of this work was to evaluate 3D general purpose PHITS MC code efficiency as an alternative approach for photon beam specification. In this study, we developed a model of an ELEKTA SL25 accelerator and used the transport code PHITS for calculating the total absorbed dose and the neutron energy spectra infield and outside the treatment field. This model was validated against measurements performed with bubble detector spectrometers and Boner sphere for 18 MV linacs, including both photons and neutrons. The average absolute difference between the calculated and measured absorbed dose for the out-of-field region was around 11%. Taking into account a simplification for simulated geometry, which does not include any potential scattering materials around, the obtained result is very satisfactorily. A good agreement between the simulated and measured neutron energy spectra was observed while comparing to data found in the literature.

  5. Measurement of cosmic-ray neutron dose onboard a polar route flight from New York to Seoul.

    PubMed

    Yasuda, Hiroshi; Lee, Jaejin; Yajima, Kazuaki; Hwang, Jung A; Sakai, Kazuo

    2011-07-01

    Exposure to cosmic radiation in operation of a jet aircraft is considered to be a part of the occupational exposure. Cosmic radiation doses received in aviation are generally evaluated by numerical model simulations. The precision of the model calculation should be verified by measurements. From the viewpoint of radiological protection, neutrons are the most contributing radiation component and have to be precisely measured. Neutron measurements were thus performed in a long-haul flight using a relatively new transportable neutron monitor (WENDI-II) which responds fairly well to the cosmic-ray neutrons. The in-flight measurement was carried out on 5-6 November 2009 on a polar route flight from New York/John F. Kennedy airport to Seoul/Incheon airport. The flying time was ~14 h. The observations obtained as 1 cm ambient dose equivalent were compared with model calculations using a computer program developed by the authors for the calculation of aviation route doses 'JISCARD EX'. Good agreements between the measured and calculated values were observed over the polar route where the geomagnetic cut-off rigidity is the lowest.

  6. An analytical model of leakage neutron equivalent dose for passively-scattered proton radiotherapy and validation with measurements.

    PubMed

    Schneider, Christopher; Newhauser, Wayne; Farah, Jad

    2015-05-18

    Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose  at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference). Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference). The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation.

  7. Relative light yield and temporal response of a stilbene-doped bibenzyl organic scintillator for neutron detection

    SciTech Connect

    Brown, J. A.; Goldblum, B. L. Brickner, N. M.; Daub, B. H.; Kaufman, G. S.; Bibber, K. van; Vujic, J.; Bernstein, L. A.; Bleuel, D. L.; Caggiano, J. A.; Hatarik, R.; Phillips, T. W.; Zaitseva, N. P.; Wender, S. A.

    2014-05-21

    The neutron time-of-flight (nTOF) diagnostics used to characterize implosions at the National Ignition Facility (NIF) has necessitated the development of novel scintillators that exhibit a rapid temporal response and high light yield. One such material, a bibenzyl-stilbene mixed single-crystal organic scintillator grown in a 99.5:0.5 ratio in solution, has become the standard scintillator used for nTOF diagnostics at NIF. The prompt fluorescence lifetime and relative light yield as a function of proton energy were determined to calibrate this material as a neutron detector. The temporal evolution of the intensity of the prompt fluorescent response was modeled using first-order reaction kinetics and the prompt fluorescence decay constant was determined to be 2.46 ± 0.01 (fit) ± 0.13 (systematic) ns. The relative response of the bibenzyl-stilbene mixed crystal generated by recoiling protons was measured, and results were analyzed using Birks' relation to quantify the non-radiative quenching of excitation energy in the scintillator.

  8. Evaluation of dose equivalent by the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities.

    PubMed

    Chau, Q; Lahaye, T

    2007-01-01

    This paper presents the results of measurements made with the electronic personal neutron Saphydose-N during the four campaigns of the European contract EVIDOS (EValuation of Individual DOSimetry in mixed neutron and photon radiation fields). These measurements were performed at Institute for Radiological Protection and Nuclear Safety (IRSN) in France (C0), at the Krümmel Nuclear Power Plant in Germany (C1), at the VENUS Research Reactor and the Belgonucléaire fuel processing plant in Belgium (C2) and at the Ringhals Nuclear Power Plant in Sweden (C3). The results for Saphydose-N are compared with reference values for dose equivalent.

  9. Determination of relative krypton fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Edwards, E. R.; Cassata, W. S.; Velsko, C. A.; Yeamans, C. B.; Shaughnessy, D. A.

    2016-11-01

    Precisely-known fission yield distributions are needed to determine a fissioning isotope and the incident neutron energy in nuclear security applications. 14 MeV neutrons from DT fusion at the National Ignition Facility induce fission in depleted uranium contained in the target assembly hohlraum. The fission yields of Kr isotopes (85m, 87, 88, and 89) are measured relative to the cumulative yield of 88Kr and compared to previously tabulated values. The results from this experiment and England and Rider are in agreement, except for the 85mKr/88Kr ratio, which may be the result of incorrect nuclear data.

  10. Cosmic-ray neutron albedo dose in low-earth orbits

    NASA Technical Reports Server (NTRS)

    Wilson, J. W.; Townsend, L. W.; Farhat, H.

    1989-01-01

    An earth albedo neutron environmental model is proposed which provides a way to estimate neutron exposure in low-earth orbit. It is shown that, in the predominantly low inclination orbits (i=28.5 deg) used in the U.S. space program, the neutron exposures are relatively low (0.7 cSv/y). The neutron exposures are more significant for polar orbital missions and even high inclination missions, such as Skylab (i=57 deg).

  11. Shielding application of perturbation theory to determine changes in neutron and gamma doses due to changes in shield layers

    NASA Technical Reports Server (NTRS)

    Fieno, D.

    1972-01-01

    Perturbation theory formulas were derived and applied to determine changes in neutron and gamma-ray doses due to changes in various radiation shield layers for fixed sources. For a given source and detector position, the perturbation method enables dose derivatives with respect to density, or equivalently thickness, for every layer to be determined from one forward and one inhomogeneous adjoint calculation. A direct determination without the perturbation approach would require two forward calculations to evaluate the dose derivative due to a change in a single layer. Hence, the perturbation method for obtaining dose derivatives requires fewer computations for design studies of multilayer shields. For an illustrative problem, a comparison was made of the fractional change in the dose per unit change in the thickness of each shield layer in a two-layer spherical configuration as calculated by perturbation theory and by successive direct calculations; excellent agreement was obtained between the two methods.

  12. Measurements of the neutron dose and energy spectrum on the International Space Station during expeditions ISS-16 to ISS-21.

    PubMed

    Smith, M B; Akatov, Yu; Andrews, H R; Arkhangelsky, V; Chernykh, I V; Ing, H; Khoshooniy, N; Lewis, B J; Machrafi, R; Nikolaev, I; Romanenko, R Y; Shurshakov, V; Thirsk, R B; Tomi, L

    2013-01-01

    As part of the international Matroshka-R and Radi-N experiments, bubble detectors have been used on board the ISS in order to characterise the neutron dose and the energy spectrum of neutrons. Experiments using bubble dosemeters inside a tissue-equivalent phantom were performed during the ISS-16, ISS-18 and ISS-19 expeditions. During the ISS-20 and ISS-21 missions, the bubble dosemeters were supplemented by a bubble-detector spectrometer, a set of six detectors that was used to determine the neutron energy spectrum at various locations inside the ISS. The temperature-compensated spectrometer set used is the first to be developed specifically for space applications and its development is described in this paper. Results of the dose measurements indicate that the dose received at two different depths inside the phantom is not significantly different, suggesting that bubble detectors worn by a person provide an accurate reading of the dose received inside the body. The energy spectra measured using the spectrometer are in good agreement with previous measurements and do not show a strong dependence on the precise location inside the station. To aid the understanding of the bubble-detector response to charged particles in the space environment, calculations have been performed using a Monte-Carlo code, together with data collected on the ISS. These calculations indicate that charged particles contribute <2% to the bubble count on the ISS, and can therefore be considered as negligible for bubble-detector measurements in space.

  13. Effects of growth medium and fertilizer rate on the yield response of soybeans exposed to chronic doses of ozone

    SciTech Connect

    Heagle, A.S.; Letchworth, M.B.; Mitchell, C.A.

    1983-01-01

    The objectives were to determine whether wide variation in fertilizer rates or type of growth medium would affect the response of soybeans, Glycine max 'Davis' exposed to chronic doses of ozone (O/sub 3/) in open-top field chambers. Responses to O/sub 3/ were compared for plants grown in the ground or in pots containing an artificial growth medium. In 1977, the yield of plants grown in pots containing soil, sand, and a mixture of perlite, peat moss, and vermiculite was greater than that of plants grown in the ground; in 1978, the reverse was true. However, the percentage yeild loss caused by O/sub 3/ was not affected by the growth medium either year. Separate tests were made for potted plants that received different levels of fertilizer. At moderate fertilizer rates, the yield response to different doses of O/sub 3/ was not significantly affected by fertilizer rate for either year. In 1978, plants with no fertilizer added were severely stunted and even relatively high doses of O/sub 3/ did not further decrease yield. The results suggest that plant response to O/sub 3/ will be fairly uniform over a range of substrate types and fertilizer rates when edaphic conditions are adequate to insure normal plant growth. 17 references, 5 figures, 2 tables.

  14. Project for measuring the neutron electromagnetic form factor in the reaction e{sup +}e{sup -} {yields} nn-bar at the VEPP-2000 collider

    SciTech Connect

    Golubev, V. B.; Serednyakov, S. I.; Skovpen, K. Yu. Usov, Yu. V.

    2009-04-15

    A project aimed at measuring the neutron electromagnetic form factor in the reaction e{sup +}e{sup -} {yields} nn-bar with the SND detector at the VEPP-2000 e{sup +}e{sup -} collider is presented. The results obtained for the time resolution of the NaI(Tl) counter using flash-ADC are reported along with estimates of the efficiency of separation of neutron-antineutron events.

  15. Assessment of organ doses from exposure to neutrons using the Monte Carlo technique and an image-based anatomical model

    NASA Astrophysics Data System (ADS)

    Bozkurt, Ahmet

    The distribution of absorbed doses in the body can be computationally determined using mathematical or tomographic representations of human anatomy. A whole- body model was developed from the color images of the National Library of Medicine's Visible Human Project® for simulating the transport of radiation in the human body. The model, called Visible Photographic Man (VIP-Man), has sixty-one organs and tissues represented in the Monte Carlo code MCNPX at 4-mm voxel resolution. Organ dose calculations from external neutron sources were carried out using VIP-man and MCNPX to determine a new set of dose conversion coefficients to be used in radiation protection. Monoenergetic neutron beams between 10-9 MeV and 10 GeV were studied under six different irradiation geometries: anterior-posterior, posterior-anterior, right lateral, left lateral, rotational and isotropic. The results for absorbed doses in twenty-four organs and the effective doses based on twelve critical organs are presented in tabular form. A comprehensive comparison of the results with those from the mathematical models show discrepancies that can be attributed to the variations in body modeling (size, location and shape of the individual organs) and the use of different nuclear datasets or models to derive the reaction cross sections, as well as the use of different transport packages for simulation radiation effects. The organ dose results based on the realistic VIP-Man body model allow the existing radiation protection dosimetry on neutrons to be re-evaluated and improved.

  16. Measured and Calculated Neutron Spectra and Dose Equivalent Rates at High Altitudes; Relevance to SST Operations and Space Research

    NASA Technical Reports Server (NTRS)

    Foelsche, T.; Mendell, R. B.; Wilson, J. W.; Adams, R. R.

    1974-01-01

    Results of the NASA Langley-New York University high-altitude radiation study are presented. Measurements of the absorbed dose rate and of secondary fast neutrons (1 to 10 MeV energy) during the years 1965 to 1971 are used to determine the maximum radiation exposure from galactic and solar cosmic rays of supersonic transport (SST) and subsonic jet occupants. The maximum dose equivalent rates that the SST crews might receive turn out to be 13 to 20 percent of the maximum permissible dose rate (MPD) for radiation workers (5 rem/yr). The exposure of passengers encountering an intense giant-energy solar particle event could exceed the MPD for the general population (0.5 rem/yr), but would be within these permissible limits if in such rare cases the transport descends to subsonic altitude; it is in general less than 12 percent of the MPD. By Monte Carlo calculations of the transport and buildup of nucleons in air for incident proton energies E of 0.02 to 10 GeV, the measured neutron spectra were extrapolated to lower and higher energies and for galactic cosmic rays were found to continue with a relatively high intensity to energies greater than 400 MeV, in a wide altitude range. This condition, together with the measured intensity profiles of fast neutrons, revealed that the biologically important fast and energetic neutrons penetrate deep into the atmosphere and contribute approximately 50 percent of the dose equivalant rates at SST and present subsonic jet altitudes.

  17. Risk of Developing Second Cancer From Neutron Dose in Proton Therapy as Function of Field Characteristics, Organ, and Patient Age

    SciTech Connect

    Zacharatou Jarlskog, Christina; Paganetti, Harald

    2008-09-01

    Purpose: To estimate the risk of a second malignancy after treatment of a primary brain cancer using passive scattered proton beam therapy. The focus was on the cancer risk caused by neutrons outside the treatment volume and the dependency on the patient's age. Methods and Materials: Organ-specific neutron-equivalent doses previously calculated for eight different proton therapy brain fields were considered. Organ-specific models were applied to assess the risk of developing solid cancers and leukemia. Results: The main contributors (>80%) to the neutron-induced risk are neutrons generated in the treatment head. Treatment volume can influence the risk by up to a factor of {approx}2. Young patients are subject to significantly greater risks than are adult patients because of the geometric differences and age dependency of the risk models. Breast cancer should be the main concern for females. For males, the risks of lung cancer, leukemia, and thyroid cancer were significant for pediatric patients. In contrast, leukemia was the leading risk for an adult. Most lifetime risks were <1% (70-Gy treatment). The only exceptions were breast, thyroid, and lung cancer for females. For female thyroid cancer, the treatment risk can exceed the baseline risk. Conclusion: The risk of developing a second malignancy from neutrons from proton beam therapy of a brain lesion is small (i.e., presumably outweighed by the therapeutic benefit) but not negligible (i.e., potentially greater than the baseline risk). The patient's age at treatment plays a major role.

  18. Calculation of the yields for the primary species formed from the radiolysis of liquid water by fast neutrons at temperatures between 25-350°C.

    PubMed

    Butarbutar, Sofia Loren; Sanguanmith, Sunuchakan; Meesungnoen, Jintana; Sunaryo, Geni Rina; Jay-Gerin, Jean-Paul

    2014-06-01

    Monte Carlo simulations were used to calculate the yields for the primary species (e(-)aq, H(•), H2, (•)OH and H2O2) formed from the radiolysis of neutral liquid water by mono-energetic 2 MeV neutrons at temperatures between 25-350°C. The 2 MeV neutron was taken as representative of a fast neutron flux in a reactor. For light water, the moderation of these neutrons generated elastically scattered recoil protons of ∼1.264, 0.465, 0.171 and 0.063 MeV, which at 25°C, had linear energy transfers (LETs) of ∼22, 43, 69 and 76 keV/μm, respectively. Neglecting the radiation effects due to oxygen ion recoils and assuming that the most significant contribution to the radiolysis came from these first four recoil protons, the fast neutron yields could be estimated as the sum of the yields for these protons after allowance was made for the appropriate weightings according to their energy. Yields were calculated at 10(-7), 10(-6) and 10(-5) s after the ionization event at all temperatures, in accordance with the time range associated with the scavenging capacities generally used for fast neutron radiolysis experiments. The results of the simulations agreed reasonably well with the experimental data, taking into account the relatively large uncertainties in the experimental measurements, the relatively small number of reported radiolysis yields, and the simplifications included in the model. Compared with data obtained for low-LET radiation ((60)Co γ rays or fast electrons), our computed yields for fast neutron radiation showed essentially similar temperature dependences over the range of temperatures studied, but with lower values for yields of free radicals and higher values for molecular yields. This general trend is a reflection of the high-LET character of fast neutrons. Although the results of the simulations were consistent with the experiment, more experimental data are required to better describe the dependence of radiolytic yields on temperature and to test

  19. Calculation of the yields for the primary species formed from the radiolysis of liquid water by fast neutrons at temperatures between 25-350°C.

    PubMed

    Butarbutar, Sofia Loren; Sanguanmith, Sunuchakan; Meesungnoen, Jintana; Sunaryo, Geni Rina; Jay-Gerin, Jean-Paul

    2014-06-01

    Monte Carlo simulations were used to calculate the yields for the primary species (e(-)aq, H(•), H2, (•)OH and H2O2) formed from the radiolysis of neutral liquid water by mono-energetic 2 MeV neutrons at temperatures between 25-350°C. The 2 MeV neutron was taken as representative of a fast neutron flux in a reactor. For light water, the moderation of these neutrons generated elastically scattered recoil protons of ∼1.264, 0.465, 0.171 and 0.063 MeV, which at 25°C, had linear energy transfers (LETs) of ∼22, 43, 69 and 76 keV/μm, respectively. Neglecting the radiation effects due to oxygen ion recoils and assuming that the most significant contribution to the radiolysis came from these first four recoil protons, the fast neutron yields could be estimated as the sum of the yields for these protons after allowance was made for the appropriate weightings according to their energy. Yields were calculated at 10(-7), 10(-6) and 10(-5) s after the ionization event at all temperatures, in accordance with the time range associated with the scavenging capacities generally used for fast neutron radiolysis experiments. The results of the simulations agreed reasonably well with the experimental data, taking into account the relatively large uncertainties in the experimental measurements, the relatively small number of reported radiolysis yields, and the simplifications included in the model. Compared with data obtained for low-LET radiation ((60)Co γ rays or fast electrons), our computed yields for fast neutron radiation showed essentially similar temperature dependences over the range of temperatures studied, but with lower values for yields of free radicals and higher values for molecular yields. This general trend is a reflection of the high-LET character of fast neutrons. Although the results of the simulations were consistent with the experiment, more experimental data are required to better describe the dependence of radiolytic yields on temperature and to test

  20. A method for measuring tissue-equivalent dose using a pin diode and activation foil in epithermal neutron beams with EN < 100 keV.

    PubMed

    Carolan, Martin G; Rosenfeld, Anatoly B

    2006-01-01

    Silicon (Si) pin diodes can be used for neutron dosimetry by observing the change in forward bias voltage caused by neutron induced displacement damage in the diode junction. Pin diode energy response depends on Si displacement damage KERMA (K(Si)). It is hypothesised that tissue-equivalent (TE) neutron dose could be expressed as a linear combination of K(Si) and foil activation terms. Monte Carlo simulations (MCNP) of parallel monoenergetic neutron beams incident on a cylindrical TE phantom were used to calculate TE dose, K(Si) and Au, Cu and Mn foil activations along the central axis of the phantom. For spectra with neutron energies <100 keV, it is possible to estimate the TE kerma based on silicon damage kerma and Cu or Mn foil measurements. More accurate estimates are possible for spectra where the maximum neutron energy does not exceed 30 keV. PMID:16644975

  1. The Irradiation Characteristics of the Kur Heavy Water Facility (ii) - Neutron and Gamma-Ray Dose Measurements with a Twin-Chamber

    NASA Astrophysics Data System (ADS)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2003-06-01

    The Heavy Water Neutron Irradiation Facility of the Kyoto University Reactor can supply neutron energy spectra from almost pure thermal to mainly epi-thermal, using a spectrum shifter and thermal neutron filters. We will report about the measurement of the neutron and gamma-ray doses using a twin-chamber. The used twin-chamber is the combination of a tissue-equivalent ionization-chamber and a graphite ionization chamber, with detecting volumes of 80 cc. From the comparisons between the chamber-measured dose rates and the nominal values, it was confirmed that the relative dependencies of the neutron and gamma-ray doses on the heavy water thickness, were almost the same, excepting the smaller heavy-water-thickness mode, such as CB-0000-F.

  2. Estimation of neutron-equivalent dose in organs of patients undergoing radiotherapy by the use of a novel online digital detector

    NASA Astrophysics Data System (ADS)

    Sánchez-Doblado, F.; Domingo, C.; Gómez, F.; Sánchez-Nieto, B.; Muñiz, J. L.; García-Fusté, M. J.; Expósito, M. R.; Barquero, R.; Hartmann, G.; Terrón, J. A.; Pena, J.; Méndez, R.; Gutiérrez, F.; Guerre, F. X.; Roselló, J.; Núñez, L.; Brualla-González, L.; Manchado, F.; Lorente, A.; Gallego, E.; Capote, R.; Planes, D.; Lagares, J. I.; González-Soto, X.; Sansaloni, F.; Colmenares, R.; Amgarou, K.; Morales, E.; Bedogni, R.; Cano, J. P.; Fernández, F.

    2012-10-01

    Neutron peripheral contamination in patients undergoing high-energy photon radiotherapy is considered as a risk factor for secondary cancer induction. Organ-specific neutron-equivalent dose estimation is therefore essential for a reasonable assessment of these associated risks. This work aimed to develop a method to estimate neutron-equivalent doses in multiple organs of radiotherapy patients. The method involved the convolution, at 16 reference points in an anthropomorphic phantom, of the normalized Monte Carlo neutron fluence energy spectra with the kerma and energy-dependent radiation weighting factor. This was then scaled with the total neutron fluence measured with passive detectors, at the same reference points, in order to obtain the equivalent doses in organs. The latter were correlated with the readings of a neutron digital detector located inside the treatment room during phantom irradiation. This digital detector, designed and developed by our group, integrates the thermal neutron fluence. The correlation model, applied to the digital detector readings during patient irradiation, enables the online estimation of neutron-equivalent doses in organs. The model takes into account the specific irradiation site, the field parameters (energy, field size, angle incidence, etc) and the installation (linac and bunker geometry). This method, which is suitable for routine clinical use, will help to systematically generate the dosimetric data essential for the improvement of current risk-estimation models.

  3. Effect of head phantom size on 10B and 1H[n,gamma]2H dose distributions for a broad field accelerator epithermal neutron source for BNCT.

    PubMed

    Gupta, N; Niemkiewicz, J; Blue, T E; Gahbauer, R; Qu, T X

    1993-01-01

    The effect of head phantom size on the 10B and 1H[n,gamma]2H dose distributions for a broad epithermal neutron radiation field generated by an accelerator-based epithermal neutron source for boron neutron capture therapy (BNCT) have been studied. Also two techniques for calculating the absorbed gamma dose from a measured gamma-ray source distribution are compared: a Monte Carlo technique, which is well accepted in the BNCT community, and a Point Kernel technique. The count-rate distribution in the central plane of three rectangular parallelopiped head water phantoms irradiated with an epithermal neutron field was measured with a boron trifluoride (BF3) detector. This epithermal neutron field was produced at the Ohio State University Van de Graaff Accelerator Facility. The 10B absorbed dose and the gamma-ray source have the same distribution in the head phantom as the BF3 count-rate distribution. The absorbed gamma dose from the measured source distribution was calculated using MCNP, a Monte Carlo code, and QAD-CGGP, a Point Kernel code. The most pronounced effect of phantom size on 10B absorbed dose was on the dose rate at the depth of maximum dose, dmax. An increase in dose rate at dmax was observed with a decrease in phantom size, the dose rate in the smallest phantom being larger by a factor of 1.4 than the dose rate in the largest phantom. Also, dmax for the phantoms shifted deeper with a decrease in phantom dimensions. The shift between the largest and the smallest phantoms was 6 mm. Finally, the smaller phantoms had lower entrance 10B dose as a percent of the dose at dmax, or better skin sparing. Our calculations for the gamma dose show that a Point Kernel technique can be used to calculate the dose distribution as accurately as a Monte Carlo technique, in much shorter computation times.

  4. SU-E-T-598: Parametric Equation for Quick and Reliable Estimate of Stray Neutron Doses in Proton Therapy and Application for Intracranial Tumor Treatments

    SciTech Connect

    Bonfrate, A; Farah, J; Sayah, R; Clairand, I; De Marzi, L; Delacroix, S; Herault, J; Lee, C; Bolch, W

    2015-06-15

    Purpose: Development of a parametric equation suitable for a daily use in routine clinic to provide estimates of stray neutron doses in proton therapy. Methods: Monte Carlo (MC) calculations using the UF-NCI 1-year-old phantom were exercised to determine the variation of stray neutron doses as a function of irradiation parameters while performing intracranial treatments. This was done by individually changing the proton beam energy, modulation width, collimator aperture and thickness, compensator thickness and the air gap size while their impact on neutron doses were put into a single equation. The variation of neutron doses with distance from the target volume was also included in it. Then, a first step consisted in establishing the fitting coefficients by using 221 learning data which were neutron absorbed doses obtained with MC simulations while a second step consisted in validating the final equation. Results: The variation of stray neutron doses with irradiation parameters were fitted with linear, polynomial, etc. model while a power-law model was used to fit the variation of stray neutron doses with the distance from the target volume. The parametric equation fitted well MC simulations while establishing fitting coefficients as the discrepancies on the estimate of neutron absorbed doses were within 10%. The discrepancy can reach ∼25% for the bladder, the farthest organ from the target volume. Finally, the validation showed results in compliance with MC calculations since the discrepancies were also within 10% for head-and-neck and thoracic organs while they can reach ∼25%, again for pelvic organs. Conclusion: The parametric equation presents promising results and will be validated for other target sites as well as other facilities to go towards a universal method.

  5. Assessment of individual organ doses in a realistic human phantom from neutron and gamma stimulated spectroscopy of the breast and liver

    PubMed Central

    Belley, Matthew D.; Segars, William Paul; Kapadia, Anuj J.

    2014-01-01

    Purpose: Understanding the radiation dose to a patient is essential when considering the use of an ionizing diagnostic imaging test for clinical diagnosis and screening. Using Monte Carlo simulations, the authors estimated the three-dimensional organ-dose distribution from neutron and gamma irradiation of the male liver, female liver, and female breasts for neutron- and gamma-stimulated spectroscopic imaging. Methods: Monte Carlo simulations were developed using the Geant4 GATE application and a voxelized XCAT human phantom. A male and a female whole body XCAT phantom was voxelized into 256 × 256 × 600 voxels (3.125 × 3.125 × 3.125 mm3). A monoenergetic rectangular beam of 5.0 MeV neutrons or 7.0 MeV photons was made incident on a 2 cm thick slice of the phantom. The beam was rotated at eight different angles around the phantom ranging from 0° to 180°. Absorbed dose was calculated for each individual organ in the body and dose volume histograms were computed to analyze the absolute and relative doses in each organ. Results: The neutron irradiations of the liver showed the highest organ dose absorption in the liver, with appreciably lower doses in other proximal organs. The dose distribution within the irradiated slice exhibited substantial attenuation with increasing depth along the beam path, attenuating to ∼15% of the maximum value at the beam exit side. The gamma irradiation of the liver imparted the highest organ dose to the stomach wall. The dose distribution from the gammas showed a region of dose buildup at the beam entrance, followed by a relatively uniform dose distribution to all of the deep tissue structures, attenuating to ∼75% of the maximum value at the beam exit side. For the breast scans, both the neutron and gamma irradiation registered maximum organ doses in the breasts, with all other organs receiving less than 1% of the breast dose. Effective doses ranged from 0.22 to 0.37 mSv for the neutron scans and 41 to 66 mSv for the gamma scans

  6. Assessment of individual organ doses in a realistic human phantom from neutron and gamma stimulated spectroscopy of the breast and liver

    SciTech Connect

    Belley, Matthew D.; Segars, William Paul; Kapadia, Anuj J.

    2014-06-15

    Purpose: Understanding the radiation dose to a patient is essential when considering the use of an ionizing diagnostic imaging test for clinical diagnosis and screening. Using Monte Carlo simulations, the authors estimated the three-dimensional organ-dose distribution from neutron and gamma irradiation of the male liver, female liver, and female breasts for neutron- and gamma-stimulated spectroscopic imaging. Methods: Monte Carlo simulations were developed using the Geant4 GATE application and a voxelized XCAT human phantom. A male and a female whole body XCAT phantom was voxelized into 256 × 256 × 600 voxels (3.125 × 3.125 × 3.125 mm{sup 3}). A monoenergetic rectangular beam of 5.0 MeV neutrons or 7.0 MeV photons was made incident on a 2 cm thick slice of the phantom. The beam was rotated at eight different angles around the phantom ranging from 0° to 180°. Absorbed dose was calculated for each individual organ in the body and dose volume histograms were computed to analyze the absolute and relative doses in each organ. Results: The neutron irradiations of the liver showed the highest organ dose absorption in the liver, with appreciably lower doses in other proximal organs. The dose distribution within the irradiated slice exhibited substantial attenuation with increasing depth along the beam path, attenuating to ∼15% of the maximum value at the beam exit side. The gamma irradiation of the liver imparted the highest organ dose to the stomach wall. The dose distribution from the gammas showed a region of dose buildup at the beam entrance, followed by a relatively uniform dose distribution to all of the deep tissue structures, attenuating to ∼75% of the maximum value at the beam exit side. For the breast scans, both the neutron and gamma irradiation registered maximum organ doses in the breasts, with all other organs receiving less than 1% of the breast dose. Effective doses ranged from 0.22 to 0.37 mSv for the neutron scans and 41 to 66 mSv for the gamma

  7. Optimal moderator materials at various proton energies considering photon dose rate after irradiation for an accelerator-driven ⁹Be(p, n) boron neutron capture therapy neutron source.

    PubMed

    Hashimoto, Y; Hiraga, F; Kiyanagi, Y

    2015-12-01

    We evaluated the accelerator beam power and the neutron-induced radioactivity of (9)Be(p, n) boron neutron capture therapy (BNCT) neutron sources having a MgF2, CaF2, or AlF3 moderator and driven by protons with energy from 8 MeV to 30 MeV. The optimal moderator materials were found to be MgF2 for proton energies less than 10 MeV because of lower required accelerator beam power and CaF2 for higher proton energies because of lower photon dose rate at the treatment position after neutron irradiation. PMID:26272165

  8. Optimal moderator materials at various proton energies considering photon dose rate after irradiation for an accelerator-driven ⁹Be(p, n) boron neutron capture therapy neutron source.

    PubMed

    Hashimoto, Y; Hiraga, F; Kiyanagi, Y

    2015-12-01

    We evaluated the accelerator beam power and the neutron-induced radioactivity of (9)Be(p, n) boron neutron capture therapy (BNCT) neutron sources having a MgF2, CaF2, or AlF3 moderator and driven by protons with energy from 8 MeV to 30 MeV. The optimal moderator materials were found to be MgF2 for proton energies less than 10 MeV because of lower required accelerator beam power and CaF2 for higher proton energies because of lower photon dose rate at the treatment position after neutron irradiation.

  9. Comparison of secondary neutron dose in proton therapy resulting from the use of a tungsten alloy MLC or a brass collimator system

    SciTech Connect

    Diffenderfer, Eric S.; Ainsley, Christopher G.; Kirk, Maura L.; McDonough, James E.; Maughan, Richard L.

    2011-11-15

    Purpose: To apply the dual ionization chamber method for mixed radiation fields to an accurate comparison of the secondary neutron dose arising from the use of a tungsten alloy multileaf collimator (MLC) as opposed to a brass collimator system for defining the shape of a therapeutic proton field. Methods: Hydrogenous and nonhydrogenous ionization chambers were constructed with large volumes to enable measurements of absorbed doses below 10{sup -4} Gy in mixed radiation fields using the dual ionization chamber method for mixed-field dosimetry. Neutron dose measurements were made with a nominal 230 MeV proton beam incident on a closed tungsten alloy MLC and a solid brass block. The chambers were cross-calibrated against a {sup 60}Co-calibrated Farmer chamber in water using a 6 MV x-ray beam and Monte Carlo simulations were performed to account for variations in ionization chamber response due to differences in secondary neutron energy spectra. Results: The neutron and combined proton plus {gamma}-ray absorbed doses are shown to be nearly equivalent downstream from either a closed tungsten alloy MLC or a solid brass block. At 10 cm downstream from the distal edge of the collimating material the neutron dose from the closed MLC was (5.3 {+-} 0.4) x 10{sup -5} Gy/Gy. The neutron dose with brass was (6.4 {+-} 0.7) x 10{sup -5} Gy/Gy. Further from the secondary neutron source, at 50 cm, the neutron doses remain close for both the MLC and brass block at (6.9 {+-} 0.6) x 10{sup -6} Gy/Gy and (6.3 {+-} 0.7) x 10{sup -6} Gy/Gy, respectively. Conclusions: The dual ionization chamber method is suitable for measuring secondary neutron doses resulting from proton irradiation. The results of measurements downstream from a closed tungsten alloy MLC and a brass block indicate that, even in an overly pessimistic worst-case scenario, secondary neutron production in a tungsten alloy MLC leads to absorbed doses that are nearly equivalent to those seen from brass collimators. Therefore

  10. An apparatus for the study of high temperature water radiolysis in a nuclear reactor: calibration of dose in a mixed neutron/gamma radiation field.

    PubMed

    Edwards, Eric J; Wilson, Paul P H; Anderson, Mark H; Mezyk, Stephen P; Pimblott, Simon M; Bartels, David M

    2007-12-01

    The cooling water of nuclear reactors undergoes radiolytic decomposition induced by gamma, fast electron, and neutron radiation in the core. To model the process, recombination reaction rates and radiolytic yields for the water radical fragments need to be measured at high temperature and pressure. Yields for the action of neutron radiation are particularly hard to determine independently because of the beta/gamma field also present in any reactor. In this paper we report the design of an apparatus intended to measure neutron radiolysis yields as a function of temperature and pressure. A new methodology for separation of neutron and beta/gamma radiolysis yields in a mixed radiation field is proposed and demonstrated.

  11. Experimental study of fusion neutron and proton yields produced by petawatt-laser-irradiated D2-3He or CD4-3He clustering gases

    NASA Astrophysics Data System (ADS)

    Bang, W.; Barbui, M.; Bonasera, A.; Quevedo, H. J.; Dyer, G.; Bernstein, A. C.; Hagel, K.; Schmidt, K.; Gaul, E.; Donovan, M. E.; Consoli, F.; De Angelis, R.; Andreoli, P.; Barbarino, M.; Kimura, S.; Mazzocco, M.; Natowitz, J. B.; Ditmire, T.

    2013-09-01

    We report on experiments in which the Texas Petawatt laser irradiated a mixture of deuterium or deuterated methane clusters and helium-3 gas, generating three types of nuclear fusion reactions: D(d,3He)n, D(d,t)p, and 3He(d,p)4He. We measured the yields of fusion neutrons and protons from these reactions and found them to agree with yields based on a simple cylindrical plasma model using known cross sections and measured plasma parameters. Within our measurement errors, the fusion products were isotropically distributed. Plasma temperatures, important for the cross sections, were determined by two independent methods: (1) deuterium ion time of flight and (2) utilizing the ratio of neutron yield to proton yield from D(d,3He)n and 3He(d,p)4He reactions, respectively. This experiment produced the highest ion temperature ever achieved with laser-irradiated deuterium clusters.

  12. Gold nanoparticles production using reactor and cyclotron based methods in assessment of (196,198)Au production yields by (197)Au neutron absorption for therapeutic purposes.

    PubMed

    Khorshidi, Abdollah

    2016-11-01

    Medical nano-gold radioisotopes is produced regularly using high-flux nuclear reactors, and an accelerator-driven neutron activator can turn out higher yield of (197)Au(n,γ)(196,198)Au reactions. Here, nano-gold production via radiative/neutron capture was investigated using irradiated Tehran Research Reactor flux and also simulated proton beam of Karaj cyclotron in Iran. (197)Au nano-solution, including 20nm shaped spherical gold and water, was irradiated under Tehran reactor flux at 2.5E+13n/cm(2)/s for (196,198)Au activity and production yield estimations. Meanwhile, the yield was examined using 30MeV proton beam of Karaj cyclotron via simulated new neutron activator containing beryllium target, bismuth moderator around the target, and also PbF2 reflector enclosed the moderator region. Transmutation in (197)Au nano-solution samples were explored at 15 and 25cm distances from the target. The neutron flux behavior inside the water and bismuth moderators was investigated for nano-gold particles transmutation. The transport of fast neutrons inside bismuth material as heavy nuclei with a lesser lethargy can be contributed in enhanced nano-gold transmutation with long duration time than the water moderator in reactor-based method. Cyclotron-driven production of βeta-emitting radioisotopes for brachytherapy applications can complete the nano-gold production technology as a safer approach as compared to the reactor-based method.

  13. Gold nanoparticles production using reactor and cyclotron based methods in assessment of (196,198)Au production yields by (197)Au neutron absorption for therapeutic purposes.

    PubMed

    Khorshidi, Abdollah

    2016-11-01

    Medical nano-gold radioisotopes is produced regularly using high-flux nuclear reactors, and an accelerator-driven neutron activator can turn out higher yield of (197)Au(n,γ)(196,198)Au reactions. Here, nano-gold production via radiative/neutron capture was investigated using irradiated Tehran Research Reactor flux and also simulated proton beam of Karaj cyclotron in Iran. (197)Au nano-solution, including 20nm shaped spherical gold and water, was irradiated under Tehran reactor flux at 2.5E+13n/cm(2)/s for (196,198)Au activity and production yield estimations. Meanwhile, the yield was examined using 30MeV proton beam of Karaj cyclotron via simulated new neutron activator containing beryllium target, bismuth moderator around the target, and also PbF2 reflector enclosed the moderator region. Transmutation in (197)Au nano-solution samples were explored at 15 and 25cm distances from the target. The neutron flux behavior inside the water and bismuth moderators was investigated for nano-gold particles transmutation. The transport of fast neutrons inside bismuth material as heavy nuclei with a lesser lethargy can be contributed in enhanced nano-gold transmutation with long duration time than the water moderator in reactor-based method. Cyclotron-driven production of βeta-emitting radioisotopes for brachytherapy applications can complete the nano-gold production technology as a safer approach as compared to the reactor-based method. PMID:27524041

  14. ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components

    SciTech Connect

    Keney, G.S.

    1981-08-01

    A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.

  15. The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities.

    PubMed

    Ghassoun, J; Senhou, N

    2012-04-01

    In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses.

  16. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors.

    PubMed

    Waugh, C J; Rosenberg, M J; Zylstra, A B; Frenje, J A; Séguin, F H; Petrasso, R D; Glebov, V Yu; Sangster, T C; Stoeckl, C

    2015-05-01

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule. PMID:26026524

  17. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

    SciTech Connect

    Waugh, C. J. Zylstra, A. B.; Frenje, J. A.; Séguin, F. H.; Petrasso, R. D.; Rosenberg, M. J.; Glebov, V. Yu.; Sangster, T. C.; Stoeckl, C.

    2015-05-15

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.

  18. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

    DOE PAGES

    Waugh, C. J.; Rosenberg, M. J.; Zylstra, A. B.; Frenje, J. A.; Seguin, F. H.; Petrasso, R. D.; Glebov, V. Yu.; Sangster, T. C.; Stoeckl, C.

    2015-05-27

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition,more » comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.« less

  19. A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

    SciTech Connect

    Waugh, C. J.; Rosenberg, M. J.; Zylstra, A. B.; Frenje, J. A.; Seguin, F. H.; Petrasso, R. D.; Glebov, V. Yu.; Sangster, T. C.; Stoeckl, C.

    2015-05-27

    Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.

  20. Low-Dose-Rate Californium-252 Neutron Intracavitary Afterloading Radiotherapy Combined With Conformal Radiotherapy for Treatment of Cervical Cancer

    SciTech Connect

    Zhang Min; Xu Hongde; Pan Songdan; Lin Shan; Yue Jianhua; Liu Jianren

    2012-07-01

    Purpose: To study the efficacy of low-dose-rate californium-252 ({sup 252}Cf) neutron intracavitary afterloading radiotherapy (RT) combined with external pelvic RT for treatment of cervical cancer. Methods and Materials: The records of 96 patients treated for cervical cancer from 2006 to 2010 were retrospectively reviewed. For patients with tumors {<=}4 cm in diameter, external beam radiation was performed (1.8 Gy/day, five times/week) until the dose reached 20 Gy, and then {sup 252}Cf neutron intracavitary afterloading RT (once/week) was begun, and the frequency of external beam radiation was changed to four times/week. For patients with tumors >4 cm, {sup 252}Cf RT was performed one to two times before whole-pelvis external beam radiation. The tumor-eliminating dose was determined by using the depth limit of 5 mm below the mucosa as the reference point. In all patients, the total dose of the external beam radiation ranged from 46.8 to 50 Gy. For {sup 252}Cf RT, the dose delivered to point A was 6 Gy/fraction, once per week, for a total of seven times, and the total dose was 42 Gy. Results: The mean {+-} SD patient age was 54.7 {+-} 13.7 years. Six patients had disease assessed at stage IB, 13 patients had stage IIA, 49 patients had stage IIB, 3 patients had stage IIIA, 24 patients had stage IIIB, and 1 patient had stage IVA. All patients obtained complete tumor regression (CR). The mean {+-} SD time to CR was 23.5 {+-} 3.4 days. Vaginal bleeding was fully controlled in 80 patients within 1 to 8 days. The mean {+-} SD follow-up period was 27.6 {+-} 12.7 months (range, 6-48 months). Five patients died due to recurrence or metastasis. The 3-year survival and disease-free recurrence rates were 89.6% and 87.5 %, respectively. Nine patients experienced mild radiation proctitis, and 4 patients developed radiocystitis. Conclusions: Low-dose-rate {sup 252}Cf neutron RT combined with external pelvic RT is effective for treating cervical cancer, with a low incidence of

  1. Feasibility of boron neutron capture therapy (BNCT) for malignant pleural mesothelioma from a viewpoint of dose distribution analysis

    SciTech Connect

    Suzuki, Minoru . E-mail: msuzuki@rri.kyoto-u.ac.jp; Sakurai, Yoshinori; Masunaga, Shinichiro; Kinashi, Yuko; Nagata, Kenji; Maruhashi, Akira; Ono, Koji

    2006-12-01

    Purpose: To investigate the feasibility of boron neutron capture therapy (BNCT) for malignant pleural mesothelioma (MPM) from a viewpoint of dose distribution analysis using Simulation Environment for Radiotherapy Applications (SERA), a currently available BNCT treatment planning system. Methods and Materials: The BNCT treatment plans were constructed for 3 patients with MPM using the SERA system, with 2 opposed anterior-posterior beams. The {sup 1}B concentrations in the tumor and normal lung in this study were assumed to be 84 and 24 ppm, respectively, and were derived from data observed in clinical trials. The maximum, mean, and minimum doses to the tumors and the normal lung were assessed for each plan. The doses delivered to 5% and 95% of the tumor volume, D{sub 05} and D{sub 95}, were adopted as the representative dose for the maximum and minimum dose, respectively. Results: When the D{sub 05} to the normal ipsilateral lung was 5 Gy-Eq, the D{sub 95} and mean doses delivered to the normal lung were 2.2-3.6 and 3.5-4.2 Gy-Eq, respectively. The mean doses delivered to the tumors were 22.4-27.2 Gy-Eq. The D{sub 05} and D{sub 95} doses to the tumors were 9.6-15.0 and 31.5-39.5 Gy-Eq, respectively. Conclusions: From a viewpoint of the dose-distribution analysis, BNCT has the possibility to be a promising treatment for MPM patients who are inoperable because of age and other medical illnesses.

  2. Neutron Dose and Sub-Kelvin Resistance of the Tardigrade: Ramazzottius Varieoranatus

    NASA Astrophysics Data System (ADS)

    Kletetschka, G.; Horikawa, D.; Parsons, A.; Bodnarik, J.; Chervenak, J.

    2010-04-01

    Tardigrades have never been exposed to neutron/gamma radiation. They were also never cooled down to temperatures less than 1 K. We will show the survival data of these conditions and discuss the survival mechanisms.

  3. Processing of DNA damage after exposure to a single dose of fission spectrum neutrons takes 40 hours to complete

    SciTech Connect

    Peak, J.G.; Peak, M.J.

    1996-11-01

    We have examined the time course over a period of days of repair of chromosomal single-strand breaks (SSB) induced by a single dose of JANUS fission-spectrum neutrons in the DNA of human P3 epithelial teratocarcinoma cells. When the cells are allowed a period of repair incubation the breaks are totally sealed by 7 hours. But then following these initial repair the DNA is dismantled as evidenced by the reappearance of SSBs. This secondary breakage is almost as extensive as that caused by the original neutron exposure, with a maximum at 16-18 hours. Finally, the DNA is rejoined, regaining its original size by 40 hours after irradiation. The secondary repair phenomenon may have an editing function, or it many represent the processing of residual damage left unrepaired during the initial rejoining of the backbone breaks.

  4. Continuous gamma and neutron irradiation at low doses can increase the number of stromal progenitor cell (CFU-F) in mouse bone marrow

    NASA Astrophysics Data System (ADS)

    Domaratskaya, E. I.; Tsetlin, V. V.; Bueverova, E. I.; Payushina, O. I.; Butorina, N. N.; Khrushchov, N. G.; Starostin, V. I.

    Experimental groups of male and female F1 (CBA × C57Bl/6) mice at the age of 3-4 months were exposed for 10 days to gamma irradiation (total dose 1.5 cGy, dose rate 0.15 cGy/day) or neutron irradiation (neutrons at average energy of 4.5 MeV at a total neutron flux ranging from 10 5 to 10 6 cm -2 and neutron flux density from 1 to 30 cm -2 s -1). These radiation doses were chosen so as to correspond to those received aboard spacecraft. [Mitrikas, V.G., Tsetlin, V.V., 2000. Radiation control onboard the MIR orbital manned station during the 22th solar cycle. Kosm. Issled. 38(2), 113-118.] Gamma irradiation stimulated the proliferation of femoral CFU-F, and their number increased by a factor of 1.5-4.5. The ectopic marrow grafts from γ-irradiated donors also increased in size. However, no changes in CFU-S proliferation rate and their number were observed. Neutron irradiation at a total absorbed dose of 2 × 10 -1 cGy (total neutron flux 2.8 × 10 7 cm -2) produced a 1.5-3-fold increase in the number of femoral CFU-F, but that of CFU-S remained unchanged. At a lower total absorbed dose 0.82 × 10 -2 cGy, total neutron flux 1.3 × 10 6 cm -2, the number of CFU-F remained at the control level. Therefore, the effect of radiation hormesis caused by neutron irradiation was observed at doses much lower than those of gamma irradiation.

  5. Effects of Very Low Dose Fast Neutrons on Cell Membrane And Secondary Protein Structure in Rat Erythrocytes.

    PubMed

    Saeed, A; Raouf, Gehan A; Nafee, Sherif S; Shaheen, Salem A; Al-Hadeethi, Y

    2015-01-01

    The effects of ionizing radiation on biological cells have been reported in several literatures. Most of them were mainly concerned with doses greater than 0.01 Gy and were also concerned with gamma rays. On the other hand, the studies on very low dose fast neutrons (VLDFN) are rare. In this study, we have investigated the effects of VLDFN on cell membrane and protein secondary structure of rat erythrocytes. Twelve female Wistar rats were irradiated with neutrons of total dose 0.009 Gy (241Am-Be, 0.2 mGy/h) and twelve others were used as control. Blood samples were taken at the 0, 4th, 8th, and 12th days postirradiation. Fourier transform infrared (FTIR) spectra of rat erythrocytes were recorded. Second derivative and curve fitting were used to analysis FTIR spectra. Hierarchical cluster analysis (HCA) was used to classify group spectra. The second derivative and curve fitting of FTIR spectra revealed that the most significant alterations in the cell membrane and protein secondary structure upon neutron irradiation were detected after 4 days postirradiation. The increase in membrane polarity, phospholipids chain length, packing, and unsaturation were noticed from the corresponding measured FTIR area ratios. This may be due to the membrane lipid peroxidation. The observed band shift in the CH2 stretching bands toward the lower frequencies may be associated with the decrease in membrane fluidity. The curve fitting of the amide I revealed an increase in the percentage area of α-helix opposing a decrease in the β-structure protein secondary structure, which may be attributed to protein denaturation. The results provide detailed insights into the VLDFN effects on erythrocytes. VLDFN can cause an oxidative stress to the irradiated erythrocytes, which appears clearly after 4 days postirradiation.

  6. Effects of Very Low Dose Fast Neutrons on Cell Membrane And Secondary Protein Structure in Rat Erythrocytes

    PubMed Central

    Nafee, Sherif S.; Shaheen, Salem A.; Al-Hadeethi, Y.

    2015-01-01

    The effects of ionizing radiation on biological cells have been reported in several literatures. Most of them were mainly concerned with doses greater than 0.01 Gy and were also concerned with gamma rays. On the other hand, the studies on very low dose fast neutrons (VLDFN) are rare. In this study, we have investigated the effects of VLDFN on cell membrane and protein secondary structure of rat erythrocytes. Twelve female Wistar rats were irradiated with neutrons of total dose 0.009 Gy (241Am-Be, 0.2 mGy/h) and twelve others were used as control. Blood samples were taken at the 0, 4th, 8th, and 12th days postirradiation. Fourier transform infrared (FTIR) spectra of rat erythrocytes were recorded. Second derivative and curve fitting were used to analysis FTIR spectra. Hierarchical cluster analysis (HCA) was used to classify group spectra. The second derivative and curve fitting of FTIR spectra revealed that the most significant alterations in the cell membrane and protein secondary structure upon neutron irradiation were detected after 4 days postirradiation. The increase in membrane polarity, phospholipids chain length, packing, and unsaturation were noticed from the corresponding measured FTIR area ratios. This may be due to the membrane lipid peroxidation. The observed band shift in the CH2 stretching bands toward the lower frequencies may be associated with the decrease in membrane fluidity. The curve fitting of the amide I revealed an increase in the percentage area of α-helix opposing a decrease in the β-structure protein secondary structure, which may be attributed to protein denaturation. The results provide detailed insights into the VLDFN effects on erythrocytes. VLDFN can cause an oxidative stress to the irradiated erythrocytes, which appears clearly after 4 days postirradiation. PMID:26436416

  7. Dose distributions in a human head phantom for neutron capture therapy using moderated neutrons from the 2.5 MeV proton-7Li reaction or from fission of 235U

    NASA Astrophysics Data System (ADS)

    Tanaka, Kenichi; Kobayashi, Tooru; Sakurai, Yoshinori; Nakagawa, Yoshinobu; Endo, Satoru; Hoshi, Masaharu

    2001-10-01

    The feasibility of neutron capture therapy (NCT) using an accelerator-based neutron source of the 7Li(p,n) reaction produced by 2.5 MeV protons was investigated by comparing the neutron beam tailored by both the Hiroshima University radiological research accelerator (HIRRAC) and the heavy water neutron irradiation facility in the Kyoto University reactor (KUR-HWNIF) from the viewpoint of the contamination dose ratios of the fast neutrons and the gamma rays. These contamination ratios to the boron dose were estimated in a water phantom of 20 cm diameter and 20 cm length to simulate a human head, with experiments by the same techniques for NCT in KUR-HWNIF and/or the simulation calculations by the Monte Carlo N-particle transport code system version 4B (MCNP-4B). It was found that the 7Li(p,n) neutrons produced by 2.5 MeV protons combined with 20, 25 or 30 cm thick D2O moderators of 20 cm diameter could make irradiation fields for NCT with depth-dose characteristics similar to those from the epithermal neutron beam at the KUR-HWNIF.

  8. Diagnostic assessment to estimate and minimize neutron dose rates received by occupationally exposed individuals at cyclotron facilities.

    PubMed

    Reina, L C; Silva, A X; Suita, J C; Souza, M I S; Facure, A; Silva, J C P; Furlanetto, J A D; Rebello, W

    2010-03-01

    Since 2003, radiopharmaceuticals for medical diagnostic purposes have been produced at the Instituto de Engenharia Nuclear, in Brazil, using two cyclotron accelerators - CV-28 and RDS111. As a result of the ever increasing production, a diagnostic assessment to reduce neutron dose rates received by occupationally exposed individuals during irradiation processes has been developed. The purpose of this work is to present this assessment, which is currently being applied to both the Fluorine and Iodine targets of CV-28 and RDS111 cyclotron accelerators.

  9. Dose imaging in a thorax phantom with lung-equivalent volume at the epithermal neutron beam of LVR-15 reactor.

    PubMed

    Gambarini, G; Vanossi, E; Bartesaghi, G; Carrara, M; Mariani, M; Negri, A; Burian, J; Viererbl, L; Klupak, V; Rejchrt, J

    2009-07-01

    A thorax phantom has been designed, consisting of PMMA and PE plates containing a cavity filled with a laboratory-made lung-substitute. Fricke-gel dosimeters have been placed in the lung-substitute volume, and the phantom has been irradiated at the epithermal column of LVR-15 reactor. Absorbed dose images have been obtained for both gamma radiation and charged particles emitted in the (10)B reactions with thermal neutrons. Measurements with thermoluminescence dosimeters (TLDs) and Monte Carlo (MC) calculations have been performed too, in order to attain inter-comparison of results.

  10. Experimental Data of Neutron Yields from Thick Targets Bombarded by 100 to 800 MeV / Nucleon Heavy Ions.

    2001-05-15

    Version 02 The recent experimental data by the authors listed above are summarized in this paper on differential neutron yields in energy and angle produced by 100, 155 and 180 MeV/nucleon He, 100, 155, 180 and 400 MeV/nucleon C, 100, 180, 400 MeV/nucleon Ne, 400MeV/nucleon Ar, Xe and Fe, 272 and 435MeV/nucleon Nb and 800 MeV/nucleon Si ions stopping in thick targets of C, Al, Cu, Pb and Nb. The paper referenced above is availablemore » on the RSICC web site. The numerical values of the data, which were used to plot figures in References 3, 4, 5, 6 and 8 of this paper, are available for download at no charge. To get access to the data, complete a RSICC registration form and order form. Both are available by clicking on "Ordering" from the RSICC web pages. You will be contacted with details about how to proceed.« less

  11. Fission Product Yields for 14 MeV Neutrons on 235U, 238U and 239Pu

    NASA Astrophysics Data System (ADS)

    Mac Innes, M.; Chadwick, M. B.; Kawano, T.

    2011-12-01

    We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on 235U, 238U and 239Pu. The results are from historical measurements made in the 1950s-1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for 235U and 238U, but our FPYs are generally higher for 239Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV 239Pu fission cross section is now known to be 15-20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.

  12. DS86 neutron dose: Monte Carlo analysis for depth profile of 152Eu activity in a large stone sample.

    PubMed

    Endo, S; Iwatani, K; Oka, T; Hoshi, M; Shizuma, K; Imanaka, T; Takada, J; Fujita, S; Hasai, H

    1999-06-01

    The depth profile of 152Eu activity induced in a large granite stone pillar by Hiroshima atomic bomb neutrons was calculated by a Monte Carlo N-Particle Transport Code (MCNP). The pillar was on the Motoyasu Bridge, located at a distance of 132 m (WSW) from the hypocenter. It was a square column with a horizontal sectional size of 82.5 cm x 82.5 cm and height of 179 cm. Twenty-one cells from the north to south surface at the central height of the column were specified for the calculation and 152Eu activities for each cell were calculated. The incident neutron spectrum was assumed to be the angular fluence data of the Dosimetry System 1986 (DS86). The angular dependence of the spectrum was taken into account by dividing the whole solid angle into twenty-six directions. The calculated depth profile of specific activity did not agree with the measured profile. A discrepancy was found in the absolute values at each depth with a mean multiplication factor of 0.58 and also in the shape of the relative profile. The results indicated that a reassessment of the neutron energy spectrum in DS86 is required for correct dose estimation.

  13. Gamma- and neutron continuous irradiations at low doses can increase stromal progenitor cell (cfu-f) number in mouse bone marrow

    NASA Astrophysics Data System (ADS)

    Domaratskaya, E.; Tsetlin, V.; Bueverova, E.; Payushina, O.; Butorina, N.; Starostin, V.

    Low doses of continuous gamma and neutron irradiation chosen in these experiments corresponded to those aboard a spacecraft (Mitricas, Tsetlin, 2000). F1 (CBAxC57Bl/6) male and female mice at the age of 3-4 months were used. The experimental groups of mice were exposed for 10 days to gamma irradiation (total dose 1.5 cGy, dose rate 0.15 cGy/day) or neutron irradiation (neutrons with energy of 4 MeV at flow in the range from 10-5 to 10-6 n/cm2, flow densities from 1 to 30 n/cm2sec). Gamma irradiation stimulated the proliferative rate of femoral CFU-F and raised their number 1,5-4,5-fold. The size of ectopic marrow transplants from gamma irradiated donors also increased. However, no changes in CFU-S proliferative rate and their number were observed. Neutron irradiation at total absorbed dose of 48x10-3 cGy (total neutron flow 2,8x106 n/cm2) produced a 3-fold increase of femoral CFU-F number, but CFU-S number remained unchanged. If total absorbed dose was lowered to 7x10-3 cGy (total neutron flow 1,3x105 n/cm2) CFU-F number remained at the control level. Therefore, the effect of radiation hormesis that caused by the neutron irradiation was observed at doses much lower than those of gamma irradiation. Supported in part by Russian Ministry of Education (projects ``Scientific Schools'' - 1629.2003.4).

  14. Effect of Very Low Dose Fast Neutrons on the DNA of Rats' Peripheral Blood Mononuclear Cells and Leukocytes.

    PubMed

    Nafee, Sherif S; Saeed, Abdu; Shaheen, Salem A; El Assouli, Sufian M; El Assouli, M-Zaki; Raouf, Gehan A

    2016-01-01

    The effect of very low dose fast neutrons on the chromatin and DNA of rats' peripheral blood mononuclear cells (PBMC) and leukocytes has been studied in the present work using Fourier transform infrared (FTIR) and single-cell gel electrophoresis (comet assay). Fourteen female Wistar rats were used; seven were irradiated with neutrons of 0.9 cGy (Am-Be, 0.02 cGy h(-1)), and seven others were used as control. Second derivative and curve fitting were used to analyze the FTIR spectra. In addition, hierarchical cluster analysis (HCA) was used to classify the group spectra. Meanwhile, the tail moment and percentage of DNA in the tail were used as indicators to sense the breaking and the level of damage in DNA. The analysis of FTIR spectra of the PBMC of the irradiated group revealed a marked increase in the area of phosphodiesters of nucleic acids and the area ratios of RNA/DNA and phosphodiesters/carbohydrates. A sharp significant increase and decrease in the areas of RNA and DNA ribose were recorded, respectively. In the irradiated group, leukocytes with different tail lengths were observed. The distributions of tail moments and the percentage of DNA in the tail of irradiated groups were heterogeneous. The mean value of the percentages of DNA in the tail at 0.5 h post-irradiation represented low-level damage in the DNA. Therefore, one can conclude that very low dose fast neutrons might cause changes in the DNA of PBMC at the submolecular level. It could cause low-level damage, double-strand break, and chromatin fragmentation of DNA of leukocytes.

  15. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for ^{235}U(n,fission) at Thermal and Fast Neutron Energies.

    PubMed

    Sonzogni, A A; McCutchan, E A; Johnson, T D; Dimitriou, P

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 ^{235}U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of ^{86}Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel. PMID:27081973

  16. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for 235U (n ,fission) at Thermal and Fast Neutron Energies

    NASA Astrophysics Data System (ADS)

    Sonzogni, A. A.; McCutchan, E. A.; Johnson, T. D.; Dimitriou, P.

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235U 235 fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  17. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for ^{235}U(n,fission) at Thermal and Fast Neutron Energies.

    PubMed

    Sonzogni, A A; McCutchan, E A; Johnson, T D; Dimitriou, P

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 ^{235}U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of ^{86}Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  18. Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit.

    PubMed

    El-Jaby, Samy; Richardson, Richard B

    2015-07-01

    Occupational exposures from ionizing radiation are currently regulated for airline travel (<20 km) and for missions to low-Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit.

  19. Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit.

    PubMed

    El-Jaby, Samy; Richardson, Richard B

    2015-07-01

    Occupational exposures from ionizing radiation are currently regulated for airline travel (<20 km) and for missions to low-Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit. PMID:26256622

  20. Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit

    NASA Astrophysics Data System (ADS)

    El-Jaby, Samy; Richardson, Richard B.

    2015-07-01

    Occupational exposures from ionizing radiation are currently regulated for airline travel (<20 km) and for missions to low-Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit.

  1. Modeling the radiolysis of supercritical water by fast neutrons: density dependence of the yields of primary species at 400°c.

    PubMed

    Butarbutar, Sofia Loren; Meesungnoen, Jintana; Guzonas, David A; Stuart, Craig R; Jay-Gerin, Jean-Paul

    2014-12-01

    A reliable understanding of radiolysis processes in supercritical water (SCW)-cooled reactors is crucial to developing chemistry control strategies that minimize the corrosion and degradation of materials. However, directly measuring the chemistry in reactor cores is difficult due to the extreme conditions of high temperature and pressure and mixed neutron and gamma-radiation fields, which are incompatible with normal chemical instrumentation. Thus, chemical models and computer simulations are an important route of investigation for predicting the detailed radiation chemistry of the coolant in a SCW reactor and the consequences for materials. Surprisingly, information on the fast neutron radiolysis of water at high temperatures is limited, and even more so for fast neutron irradiation of SCW. In this work, Monte Carlo simulations were used to predict the G values for the primary species e(-)aq, H(•), H2, (•)OH and H2O2 formed from the radiolysis of pure, deaerated SCW (H2O) by 2 MeV monoenergetic neutrons at 400°C as a function of water density in the range of ∼0.15-0.6 g/cm(3). The 2 MeV neutron was taken as representative of a fast neutron flux in a reactor. For light water, the moderation of these neutrons after knock-on collisions with water molecules generated mostly recoil protons of 1.264, 0.465, 0.171 and 0.063 MeV. Neglecting oxygen ion recoils and assuming that the most significant contribution to the radiolysis came from these first four recoil protons, the fast neutron yields were estimated as the sum of the G values for these protons after appropriate weightings were applied according to their energy. Calculated yields were compared with available experimental data and with data obtained for low-LET radiation. Most interestingly, the reaction of H(•) atoms with water was found to play a critical role in the formation yields of H2 and (•)OH at 400°C. Recent work has underscored the potential importance of this reaction above 200°C, but its

  2. SU-E-T-594: Out-Of-Field Neutron and Gamma Dose Estimated Using TLD-600/700 Pairs in the Wobbling Proton Therapy System

    SciTech Connect

    Chen, Y; Lin, Y; Tsai, H

    2015-06-15

    Purpose: Secondary fast neutrons and gamma rays are mainly produced due to the interaction of the primary proton beam with the beam delivery nozzle. These secondary radiation dose to patients and radiation workers are unwanted. The purpose of this study is to estimate the neutron and gamma dose equivalent out of the treatment volume during the wobbling proton therapy system. Methods: Two types of thermoluminescent (TL) dosimeters, TLD-600 ({sup 6}LiF: Mg, Ti) and TLD-700 ({sup 7}LiF: Mg, Ti) were used in this study. They were calibrated in the standard neutron and gamma sources at National Standards Laboratory. Annealing procedure is 400°C for 1 hour, 100°C for 2 hours and spontaneously cooling down to the room temperature in a programmable oven. Two-peak method (a kind of glow curve analysis technique) was used to evaluate the TL response corresponding to the neutron and gamma dose. The TLD pairs were placed outside the treatment field at the neutron-gamma mixed field with 190-MeV proton beam produced by the wobbling system through the polyethylene plate phantom. The results of TLD measurement were compared to the Monte Carlo simulation. Results: The initial experiment results of calculated dose equivalents are 0.63, 0.38, 0.21 and 0.13 mSv per Gy outside the field at the distance of 50, 100, 150 and 200 cm. Conclusion: The TLD-600 and TLD-700 pairs are convenient to estimate neutron and gamma dosimetry during proton therapy. However, an accurate and suitable glow curve analysis technique is necessary. During the wobbling system proton therapy, our results showed that the neutron and gamma doses outside the treatment field are noticeable. This study was supported by the grants from the Chang Gung Memorial Hospital (CMRPD1C0682)

  3. M-BAND Analysis of Chromosome Aberration In Human Epithelial Cells exposed to Gamma-ray and Secondary Neutrons of Low Dose Rate

    NASA Technical Reports Server (NTRS)

    Hada, M.; Saganti, P. B.; Gersey, B.; Wilkins, R.; Cucinotta, F. A.; Wu, H.

    2007-01-01

    High-energy secondary neutrons, produced by the interaction of galactic cosmic rays with the atmosphere, spacecraft structure and planetary surfaces, contribute to a significant fraction to the dose equivalent in crew members and passengers during commercial aviation travel, and astronauts in space missions. The Los Alamos Nuclear Science Center (LANSCE) neutron facility's "30L" beam line is known to generate neutrons that simulate the secondary neutron spectrum of the Earth's atmosphere at high altitude. The neutron spectrum is also similar to that measured onboard spacecraft like the MIR and the International Space Station (ISS). To evaluate the biological damage, we exposed human epithelial cells in vitro to the LANSCE neutron beams at an entrance dose rate of 2.5 cGy/hr or gamma-ray at 1.7cGy/hr, and assessed the induction of chromosome aberrations that were identified with mBAND. With this technique, individually painted chromosomal bands on one chromosome allowed the identification of inter-chromosomal aberrations (translocation to unpainted chromosomes) and intra-chromosomal aberrations (inversions and deletions within a single painted chromosome). Compared to our previous results for gamma-rays and 600 MeV/nucleon Fe ions of high dose rate, the neutron data showed a higher frequency of chromosome aberrations. However, detailed analysis of the inversion type revealed that all of the three radiation types in the study induced a low incidence of simple inversions. The low dose rate gamma-rays induced a lower frequency of chromosome aberrations than high dose rate gamma-rays, but the inversion spectrum was similar for the same cytotoxic effect. The distribution of damage sites on chromosome 3 for different radiation types will also be discussed.

  4. Measurement and model prediction of proton-recoil track length distributions in NTA film dosimeters for neutron energy spectroscopy and retrospective dose assessment

    NASA Astrophysics Data System (ADS)

    Taulbee, Timothy D.

    The goal of this research was to determine whether neutron dose reconstruction could be improved through re-analysis of historic NTA films worn by workers in the 1950 through the 1970s. To improve neutron dose reconstruction, the underlying neutron energy spectra is critical in determining the organ dose due to energy dependence of the dose conversion factor as well as the application of radiation weighting factors used in epidemiology and probability of causation calculations. Monte Carlo models of proton-recoil track length distributions were developed and benchmarked against measurement data for both NTA and Ilford films. These models, when applied to several NTA film dosimeter configurations, demonstrated that proton-recoil track length distributions change based upon incident neutron energy. The neutron energy spectra changes that result from the general work environment such as source term and shielding can subsequently be modeled to predict the response of the NTA film dosimeter. An Automatic NTA Film Analyzer has been designed and developed to determine if the difference in proton-recoil track length distributions predicted by the Monte Carlo models could be measured and whether these differences could be correlated to the incident neutron energy spectra. The design required the development of a 2D-3D hybrid track recognition algorithm for a three dimensional analysis of the NTA film in order to accurately determine the proton-recoil track length for subsequent neutron energy determination. NTA films exposed to a plutonium fluoride (PuF4) and polonium boron (PoB) calibration sources were measured and compared. The proton-recoil track lengths were used to reconstruct the incident neutron energy spectra demonstrating the functionality of the analyzer and that reconstruction of the neutron energy spectra from NTA films is feasible. These measurements were compared to the Monte Carlo models and confirmed the applicability of using models to determine the NTA

  5. Effect of pulsed dose in simultaneous and sequential irradiation of V-79 cells by 14. 8-MeV neutrons and /sup 60/Co photons

    SciTech Connect

    Higgins, P.D.; DeLuca, P.M. Jr.; Gould, M.N.

    1984-09-01

    The effect of irradiating V-79 Chinese hamster cells with a mixture of 40% 14.8-MeV neutrons and 60% /sup 69/Co photons with simultaneous or sequential exposures is investigated. Sample doses are obtained by irradiating cells with alternating 3-min pulses of neutrons and photons (in the sequential case) or with mixed neutrons and photons followed by equal beam-off periods to ensure equal total exposure times for sequential and simultaneous irradiations. Differences between the survival results under each beam configuration that are consistent with previous observations with nonpulsed irradiations are observed.

  6. Effect of pulsed dose in simultaneous and sequential irradiation of V-79 cells by 14. 8 MeV neutrons and /sup 60/Co photons

    SciTech Connect

    Higgins, P.D.; DeLuca, P.M. Jr.; Gould, M.N.; Schell, M.C.; Pearson, D.W.

    1983-01-01

    The effect of irradiating V-79 Chinese hamster ovary cells with a mixture of 40% 14.8-MeV neutrons and 60% /sup 60/Co photons with simultaneous or sequential exposures is investigated. Target doses are obtained by irradiating cell samples with 3-minute-long pulses of alternating neutrons and photons (in the sequential case) or with mixed neutrons and photons followed by equal beam-off periods to insure equal total-exposure times for sequenced and simultaneous irradiations. We observe qualitative differences between the survival results under each beam configuration that confirms earlier observations.

  7. Comparison of neutron dose quantities and instrument and dosemeter readings at representative locations in an MOX fuel fabrication plant

    NASA Astrophysics Data System (ADS)

    Bartlett, D. T.; Hager, L. G.; Tanner, R. J.; Haley, R. M.; Cooper, A. J.

    2002-01-01

    The relationships between operational and protection quantities, and values of personal dosemeter and instrument readings have been determined for a recently designed MOX fuel fabrication plant. The relationships between the quantities, and the readings of personal dosemeters are sensitive to both the energy and direction distribution of neutron fluence. The energy distributions were calculated using the Monte Carlo code MCBEND. The direction distribution was addressed by calculating independently, spectral components for which the direction distribution could be reasonably assumed. At representative locations, and for assumed worker orientations, the radiation field is analysed as having, in general, three components—a direct, unidirectional component from the nearest identified discrete source, which is considered incident A-P, several unidirectional components from other such sources which are treated as a rotational component and a scattered isotropic component. The calculated spectra were folded with conversion coefficients for personal dose equivalent, Hp(10) slab (A-P, ROT and ISO), effective dose, E, (A-P, ROT and ISO), ambient dose equivalent, H*(10), personal dosemeter (AP, ROT and ISO) and survey instrument response characteristics.

  8. Fluence-to-dose conversion coefficients from monoenergetic neutrons below 20 MeV based on the VIP-Man anatomical model

    NASA Astrophysics Data System (ADS)

    Bozkurt, A.; Chao, T. C.; Xu, X. G.; Bozkurt, A.; Chao, T. C.

    2000-10-01

    A new set of fluence-to-absorbed dose and fluence-to-effective dose conversion coefficients have been calculated for neutrons below 20 MeV using a whole-body anatomical model, VIP-Man, developed from the high-resolution transverse colour photographic images of the National Library of Medicine's Visible Human Project®. Organ dose calculations were performed using the Monte Carlo code MCNP for 20 monoenergetic neutron beams between 1×10-9 MeV and 20 MeV under six different irradiation geometries: anterior-posterior, posterior-anterior, right lateral, left lateral, rotational and isotropic. The absorbed dose for 24 major organs and effective dose results based on the realistic VIP-Man are presented and compared with those based on the simplified MIRD-based phantoms reported in the literature. Effective doses from VIP-Man are not significantly different from earlier results for neutrons in the energy range studied. There are, however, remarkable deviations in organ doses due to the anatomical differences between the image-based and the earlier mathematical models.

  9. Measurements of the neutron yields from 7Li(p,n)7Be reaction (thick target) with incident energies from 1.885 to 2.0 MeV.

    PubMed

    Yu, W; Yue, G; Han, X; Chen, J; Tian, B

    1998-07-01

    Accelerator-based neutron source have been considered to be practical for boron neutron capture therapy (BNCT). Based on experience with a parameters of the Brookhaven National Laboratory BMRR reactor neutron source, which has been used in treatment experiments, the future accelerator-based neutron source for BNCT should have the properties of low energy distribution (< 100 keV) and high flux (about 10(9) neutrons per second per square centimeter) in the patient zone. Using protons to bombard thick 7Li targets, generating neutrons via the 7Li(p,n)7Be reaction, is one of the optimal choices for this kind of neutron source. Neutron yield data versus incident energy are necessary in order to select the proper incident energy and for estimating how high the incident proton current should be. The required proton beam current intensity is one of the key parameters for an accelerator useful for BNCT. In the present work, neutron yields of the 7Li(p,n)7Be reaction with a thick lithium target and incident energies of 1.885 and 1.9 MeV were measured at 0 degree with respect to the incident beam direction. The results are (3.08 +/- 0.17) x 10(12) and (5.71 +/- 0.32) x 10(12) neutrons/C sr, respectively. Neutron yield angular distribution measurements at 2 MeV incident energy were also performed. The proton beams were generated by the Peking University 4.5 MV electrostatic accelerator. The emitted neutrons from these reactions have the advantages of low energy distribution and forward angular distribution, which are requirements for a BNCT neutron source. The data obtained in this work can be used as a reference to study the accelerator-based neutron sources for BNCT.

  10. RBE values and repair characteristics for colo-rectal injury after caesium 137 gamma-ray and neutron irradiation. II. Fractionation up to ten doses.

    PubMed

    Terry, N H; Denekamp, J

    1984-07-01

    Early and late colo-rectal damage in mice have been assessed after 137Cs gamma irradiation and 3 MeV neutrons given as 1,2,5 or 10 fractions. Damage was measured by early changes in body weight, the late production of short faecal pellets and the pattern of lethality after irradiation. The data have been analysed in terms of the time course of expression of damage, fractionation effects and the RBE for neutrons over a wide range of doses per fraction (0.5-12.5 Gy neutrons, 3.5-33.5 Gy gamma rays). An initial epithelial denudation led to an early loss of weight, maximal at 11-17 days after irradiation. A dose-dependent weight reduction persisted over the animals' life-time. Deaths after localised pelvic gamma irradiation were progressive with no sharp demarcation between early or late phases of injury. The time course for lethality was qualitatively similar after neutrons. Beyond six months the rectum became constricted by fibrosis and a higher proportion of small faecal pellets was observed. At 6-15 months relatively shallow dose-response curves were obtained for this change. The sparing effect of fractionation was marked for the gamma-irradiated mice and almost absent after neutrons. A very high repair increment (11 Gy) was seen with two gamma-ray fractions of 20 Gy. At lower doses per fraction the proportion of each gamma-ray fraction recovered was 50-69% for all assays, i.e., similar to that for other normal tissues. There was a slight enhancement in the sparing effect for the late compared with the early assays over the lower dose range. The RBE was strongly dependent on dose per fraction because of the lack of reparable damage after neutrons. The RBE for both early and late effects was 5.0 at a neutron dose per fraction of 1 Gy. Extrapolation of the RBE data to lower doses, using the linear quadratic model, predicts a higher RBE for late (7.4-12.7) than for early damage (5.7-8.5) if gamma-ray doses below 5 Gy are used.

  11. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

    PubMed

    Schmitz, Tobias; Blaickner, Matthias; Schütz, Christian; Wiehl, Norbert; Kratz, Jens V; Bassler, Niels; Holzscheiter, Michael H; Palmans, Hugo; Sharpe, Peter; Otto, Gerd; Hampel, Gabriele

    2010-10-01

    To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also

  12. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

    PubMed

    Schmitz, Tobias; Blaickner, Matthias; Schütz, Christian; Wiehl, Norbert; Kratz, Jens V; Bassler, Niels; Holzscheiter, Michael H; Palmans, Hugo; Sharpe, Peter; Otto, Gerd; Hampel, Gabriele

    2010-10-01

    To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also

  13. Low cycle fatigue properties of reduced activation ferritic/martensitic steels after high-dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Aktaa, J.; Povstyanko, A.; Prokhorov, V.; Diegele, E.; Lässer, R.

    2011-08-01

    This paper focuses on the low cycle fatigue (LCF) behaviour of reduced activation ferritic/martensitic steels irradiated to a displacement damage dose of up to 70 dpa at 330-337 °C in the BOR 60 reactor within the ARBOR 2 irradiation programme. The influence of neutron irradiation on the fatigue behaviour was determined for the as-received EUROFER97, pre-irradiation heat-treated EUROFER97 HT and F82H-mod steels. Strain-controlled push-pull loading was performed using miniaturized cylindrical specimens at a constant temperature of 330 °C with total strain ranges between 0.8% and 1.1%. Comparison of the LCF behaviour of irradiated and reference unirradiated specimens was performed for both the adequate total and inelastic strains. Neutron irradiation-induced hardening may have various effects on the fatigue behaviour of the steels. The reduction of inelastic strain in the irradiated state compared with the reference unirradiated state at common total strain amplitudes may increase fatigue lifetime. The increase in the stress at the adequate inelastic strain, by contrast, may accelerate fatigue damage accumulation. Depending on which of the two effects mentioned dominates, neutron irradiation may either extend or reduce the fatigue lifetime compared with the reference unirradiated state. The results obtained for EUROFER97 and EUROFER97 HT confirm these considerations. Most of the irradiated specimens show fatigue lifetimes comparable to those of the reference unirradiated state at adequate inelastic strains. Some irradiated specimens, however, show lifetime reduction or increase in comparison with the reference state at adequate inelastic strains.

  14. Ion dose dependence of the sputtering yield: Ar{sup +}, Ne{sup +}, and Xe{sup +} bombardment of Ru(0001) and Al(111)

    SciTech Connect

    Burnett, J.W.; Pellin, M.J.; Whitten, J.E.; Gruen, D.M.; Yates, J.T. Jr.

    1994-04-01

    The sputtering yield from clean metal surfaces has long been considered to be insensitive to primary ion dose at moderate ion fluences (< 10{sup 18} ions/cm{sup 2}). Using carefully cleaned and well-characterized targets, the ion dose dependence of the sputtering yield of Ru(0001) and Al(111) has been investigated. The sputtering yield of Ru(0001) is found to decrease substantially following primary ion bombardment at low fluences, while the sputtering yield of Al(111) exhibits no fluence dependence at low primary ion dose. Using secondary neutral mass spectrometry (SNMS), the sputtering yield of ruthenium was observed to decrease following ion bombardment by argon, xenon, and neon. High-detection-efficiency time-of-flight mass spectrometry was coupled with nonresonant laser ionization to allow real-time sputtering yield measurements and to minimize target damage during data collection. The experiments show that the sputtering yield of Ru(0001) decreases by 50%, following a primary ion fluence of, less than 10{sup 16} ions/cm{sup 2} for sputtering by either argon or neon ions and by 25%, following primary ion fluences of less than 10{sup 14} ions/cm{sup 2} for sputtering by xenon. The small size of the experimentally determined damage cross section suggests that microscopic changes in the surface structure cause the observed sputtering yield depression. In contrast to the ruthenium results, the sputtering yield of Al(111) appears to be insensitive to primary ion fluence at low fluences. Calculations using the TRansport of Ions in Matter (TRIM) Monte Carlo sputtering simulation were carried out to investigate the effect of primary ion implantation upon the sputtering yield of ruthenium as well as the effect of a reduced surface binding energy of ruthenium surface atoms. The TRIM results indicate that neither of these mechanisms can explain the experimentally observed fluence dependence of the sputtering yield of ruthenium.

  15. Predicted neutron yield and radioactivity for laser-induced (p,n) reactions in LiF

    SciTech Connect

    Swift, D C; McNaney, J M

    2009-01-30

    Design calculations are presented for a pulsed neutron source comprising polychromatic protons accelerated from a metal foil by a short-pulse laser, and a LiF converter in which (p,n) reactions occur. Although the proton pulse is directional, neutrons are predicted to be emitted relatively isotropically. The neutron spectrum was predicted to be similar to the proton spectrum, but with more neutrons of low energy in the opposite direction to the incident protons. The angular dependence of spectrum and intensity was predicted. The (p,n) reactions generate unstable nuclei which decay predominantly by positron emission to the original {sup 7}Li and {sup 19}F isotopes. For the initial planned experiments using a converter 1mm thick, we predict that 0.1% of the protons will undergo a (p,n) reaction, producing 10{sup 9} neutrons. Ignoring the unreacted protons, neutrons, and prompt gamma emission as excited nuclear states decay, residual positron radioactivity (and production of pairs of 511 keV annihilation photons) is initially 4.2MBq decaying with a half-life of 17.22 s for 6 mins ({sup 19}Ne decays), then 135Bq decaying with a half-life of 53.22 days ({sup 7}Be decays).

  16. Processing of DNA damage after exposure to a single dose of fission spectrum neutrons takes 40 hours to complete

    SciTech Connect

    Peak, J.G.; Peak, M.J.

    1994-01-01

    The authors have examined the long-term (days) fate of breaks induced in the DNA of human P3 epithelial teratocarcinoma cells by a single dose of JANUS fission-spectrum neutrons (mean energy 0.85 MeV). We used alkaline-filter elution methods that assay totality of single- and double-strand breaks, generally referred to as single-strand breaks (SSBs). When the cells are allowed a period of repair incubation, these breaks are totally sealed by 7 hours after the original exposure, but following the initial repair the DNA is dismantled, as revealed by the reappearance of SSBS. This secondary breakage is almost as extensive as that caused by the original neutron exposure, with a maximum at 16-18 hours after irradiation. Finally, the DNA is once again rejoined, regaining its original size by 40 hours after irradiation. The secondary repair phenomenon may have an editing function, or it may represent the processing of residual damage left unrepaired during the initial rejoining of the backbone breaks.

  17. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

    SciTech Connect

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  18. High-dose neutron irradiation of Hi-Nicalon Type S silicon carbide composites. Part 2: Mechanical and physical properties

    NASA Astrophysics Data System (ADS)

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wally; Snead, Lance L.

    2015-07-01

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573-1073 K. The material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. The observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  19. Low-dose ticagrelor yields an antiplatelet efficacy similar to that of standard-dose ticagrelor in healthy subjects: an open-label randomized controlled trial.

    PubMed

    Li, Pan; Gu, Ying; Yang, Yawei; Chen, Lizhi; Liu, Junmei; Gao, Lihong; Qin, Yongwen; Cai, Quancai; Zhao, Xianxian; Wang, Zhuo; Ma, Liping

    2016-01-01

    Ticagrelor has a greater antiplatelet efficacy than clopidogrel but may be accompanied by an increased risk of bleeding. This study evaluated the antiplatelet effect and pharmacokinetic profile of low-dose ticagrelor in healthy Chinese volunteers. Thirty healthy subjects were randomized to receive standard-dose ticagrelor (180-mg loading dose, 90-mg twice daily [bid] [n = 10]), low-dose ticagrelor (90-mg loading dose, 45-mg bid [n = 10]), or clopidogrel (600-mg loading dose, 75-mg once daily [n = 10]). Platelet reactivity was assessed by using the VerifyNow P2Y12 assay at baseline and 0.5, 1, 2, 4, 8, 24, 48, and 72 hours post-dosing. The ticagrelor and AR-C124910XX concentrations were measured for pharmacokinetic analysis. The percentage inhibition of P2Y12 reaction units was higher in the low-dose and standard-dose ticagrelor group than in the clopidogrel group at 0.5, 1, 2, 4, 8, and 48 hours post-dosing (P < 0.05 for all), but did not differ significantly between the two ticagrelor doses at any time-point (P > 0.05). The plasma ticagrelor and ARC124910XX concentrations were approximately 2-fold higher with standard-dose versus low-dose ticagrelor. No serious adverse events were reported. In conclusion, low-dose ticagrelor achieved faster and higher inhibition of platelet functions in healthy Chinese subjects than did clopidogrel, with an antiplatelet efficacy similar to that of standard-dose ticagrelor. PMID:27554803

  20. Low-dose ticagrelor yields an antiplatelet efficacy similar to that of standard-dose ticagrelor in healthy subjects: an open-label randomized controlled trial

    PubMed Central

    Li, Pan; Gu, Ying; Yang, Yawei; Chen, Lizhi; Liu, Junmei; Gao, Lihong; Qin, Yongwen; Cai, Quancai; Zhao, Xianxian; Wang, Zhuo; Ma, Liping

    2016-01-01

    Ticagrelor has a greater antiplatelet efficacy than clopidogrel but may be accompanied by an increased risk of bleeding. This study evaluated the antiplatelet effect and pharmacokinetic profile of low-dose ticagrelor in healthy Chinese volunteers. Thirty healthy subjects were randomized to receive standard-dose ticagrelor (180-mg loading dose, 90-mg twice daily [bid] [n = 10]), low-dose ticagrelor (90-mg loading dose, 45-mg bid [n = 10]), or clopidogrel (600-mg loading dose, 75-mg once daily [n = 10]). Platelet reactivity was assessed by using the VerifyNow P2Y12 assay at baseline and 0.5, 1, 2, 4, 8, 24, 48, and 72 hours post-dosing. The ticagrelor and AR-C124910XX concentrations were measured for pharmacokinetic analysis. The percentage inhibition of P2Y12 reaction units was higher in the low-dose and standard-dose ticagrelor group than in the clopidogrel group at 0.5, 1, 2, 4, 8, and 48 hours post-dosing (P < 0.05 for all), but did not differ significantly between the two ticagrelor doses at any time-point (P > 0.05). The plasma ticagrelor and ARC124910XX concentrations were approximately 2-fold higher with standard-dose versus low-dose ticagrelor. No serious adverse events were reported. In conclusion, low-dose ticagrelor achieved faster and higher inhibition of platelet functions in healthy Chinese subjects than did clopidogrel, with an antiplatelet efficacy similar to that of standard-dose ticagrelor. PMID:27554803

  1. Direction distributions of neutrons and reference values of the personal dose equivalent in workplace fields.

    PubMed

    Luszik-Bhadra, M; Bolognese-Milsztajn, T; Boschung, M; Coeck, M; Curzio, G; d'Errico, F; Fiechtner, A; Lacoste, V; Lindborg, L; Reginatto, M; Schuhmacher, H; Tanner, R; Vanhavere, F

    2007-01-01

    Within the EC project EVIDOS, double-differential (energy and direction) fluence spectra were determined by means of novel direction spectrometers. By folding the spectra with fluence-to-dose equivalent conversion coefficients, contributions to H*(10) for 14 directions, and values of the personal dose equivalent Hp(10) and the effective dose E for 6 directions of a person's orientation in the field were determined. The results of the measurements and calculations obtained within the EVIDOS project in workplace fields in nuclear installations in Europe, i.e., at Krümmel (boiling water reactor and transport cask), at Mol (Venus research reactor and fuel facility Belgonucléaire) and at Ringhals (pressurised reactor and transport cask) are presented. PMID:17369265

  2. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    NASA Astrophysics Data System (ADS)

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2015-01-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50-70 °C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5 × 1025 m-2 to reach the total ion fluence of 1 × 1026 m-2 in order to investigate the near-surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate the irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near-surface (<5 µm depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near-surface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.3 dpa) at 500 °C cases even in the relatively low ion fluence of 1026 m-2.

  3. The ratio R{sub dp} of the quasielastic nd {yields} p(nn) to the elastic np {yields} pn charge-exchange-process yields at the proton emitting angle {theta}{sub p,lab} = 0 deg. over 0.55-2.0 GeV neutron beam energy region. Experimental results

    SciTech Connect

    Sharov, V. I. Morozov, A. A.; Shindin, R. A.; Antonenko, V. G.; Borzakov, S. B.; Borzunov, Yu. T.; Chernykh, E. V.; Chumakov, V. F.; Dolgii, S. A.; Finger, M.; Finger, M.; Golovanov, L. B.; Guriev, D. K.; Janata, A.; Kirillov, A. D.; Kovalenko, A. D.; Krasnov, V. A.; Kuzmin, N. A.; Kurilkin, A. K.; Kurilkin, P. K.

    2009-06-15

    New experimental results on ratio R{sub dp} of the quasielastic charge-exchange yield at the outgoing proton angle {theta}{sub p,lab} = 0 deg. for the nd {yields} p(nn) reaction to the elastic np {yields} pn charge-exchange yield, are presented. The measurements were carried out at the Nuclotron of the Veksler and Baldin Laboratory of High Energies of the JINR (Dubna) at the neutron-beam kinetic energies of 0.55, 0.8, 1.0, 1.2, 1.4, 1.8, and 2.0 GeV. The intense neutron beam with small momentum spread was produced by breakup of deuterons which were accelerated and extracted to the experimental hall. In both reactions mentioned above the outgoing protons with the momenta p{sub p} approximately equal to the neutron-beam momentum p{sub n,beam} were detected in the directions close to the direction of incident neutrons, i.e., in the vicinity of the scattering angle {theta}{sub p,lab} = 0 deg. Measured in the same data-taking runs, the angular distributions of the charge-exchange-reaction products were corrected for the well-known instrumental effects and averaged in the vicinity of the incident-neutron-beam direction. These corrected angular distributions for every of nd {yields} p(nn) and np {yields} pn charge-exchange processes were proportional to the differential cross sections of the corresponding reactions. The data were accumulated by Delta-Sigma setup magnetic spectrometer with two sets of multiwire proportional chambers located upstream and downstream of the momentum analyzing magnet. Inelastic processes were considerably reduced by the additional detectors surrounding the hydrogen and deuterium targets. The time-of-flight system was applied to identify the detected particles. The accumulated data treatment and analysis, as well as possible sources of the systematic errors are discussed.

  4. Evaluating a Contribution of the Knock-on Deuterons to the Neutron Yield in the Experiments with Weakly Collisional Plasma Jets (Part 1)

    SciTech Connect

    Ryutov, D. D.

    2015-12-01

    Laser-generated interpenetrating plasma jets are widely used in the studies of collisionless interaction of counter-streaming plasmas in conjunction with possible formation of collisionless shocks. In a number of experiments of this type the plasma is formed on plastic targets made of CH or CD. The study of the DD neutron production from the interaction between two CD jets on the one hand and between a CD jet and a CH jet could serve as a qualitative indicator of the collisionless shock formation. The purpose of this memo is a discussion of the effect of collisions on the neutron generation in the interpenetrating CH and CD jets. First, the kinematics of the large-deflection collisions of the deuterons and carbon are discussed. Then the scattering angles are related with the corresponding Rutherford cross-section. After that expression for the number of the backscattered deuterons is provided, and their contribution to the neutron yield is evaluated. The results may be of some significance to the kinetic codes benchmarking and developing the neutron diagnostic.

  5. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Vieira, Dave; Wilhelmy, Jerry

    2014-03-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for certain fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He,or3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers, relative fission product yield were determined at a high precision of 2-3 % as well as absolute fission product yields at a lower precision of 5-6 %. Preliminary results will be presented for a number of fission product isotopes over the incident neutron energy range of 0.5 to 14.8 MeV.

  6. Evaluating a Contribution of the Knock-on Deuterons to the Neutron Yield in the Experiments with Weakly Collisional Plasma Jets (Part 2)

    SciTech Connect

    Ryutov, D. D.

    2015-12-08

    Part 1 of this note considered the kinematics of large-angle scattering (LAS) of the deuterons on the counter-streaming carbon ions, with both flows having the same velocity V. Due to a large mass ratio mC/mD, the backscattered deuterons have high velocity of up to (24/7)V. This significantly increases the cross-section for the neutron production in the collisions between the back-scattered and incoming deuterons and may provide significant contribution to the total neutron yield, despite the smallness of a large-angle Coulomb cross-section. This effect becomes particularly important when only one of the colliding streams is made of CD, whereas the other stream is made of CH. Part 1 evaluated the neutron yield produced by this mechanism and have found that its relative role increases for higher plasma densities and lower velocities. Part 2 discusses signatures of this effect which can be used to identify it experimentally and also discusses in some more detail its spatio-temporal characteristics. It goes without saying that a complete quantitative assessment should be based on numerical simulations accounting for the large-angle scattering.

  7. Neutron tube design study for boron neutron capture therapy application

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1999-05-06

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  8. High dose effects in neutron irradiated face-centered cubic metals

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors.

  9. Application of LiF thermoluminescence dosimeter powders in neutron gamma mixed field dosimetry and dose mapping in the thermal column of a triga Mk II reactor

    NASA Astrophysics Data System (ADS)

    Mukherjee, Bhaskar; Böck, Helmut; Vana, Norbert

    1987-02-01

    Thermal neutron and fission product gamma dose rates at different positions in the horizontal thermal column of a TRIGA Mk-II reactor, as well as in a cadmium cladded cylindrical cavity embedded in a borated wooden box placed in the thermal column, were assessed with TLD-600 and TLD-700 dosimeter powders.

  10. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    SciTech Connect

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².

  11. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    DOE PAGES

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Finalmore » thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².« less

  12. Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

    SciTech Connect

    Field, Kevin G.; Yang, Ying; Allen, Todd R.; Busby, Jeremy T.

    2015-05-01

    Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling to mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were utilized to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and includes expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibit unique defect sink characteristics depending on their local structure. Such interactions were found to be consistent across all doses investigated and had larger global implications including precipitation of Ni-Si clusters near different grain boundary types.

  13. Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

    SciTech Connect

    Field, Kevin G.; Yang, Ying; Busby, Jeremy T.; Allen, Todd R.

    2015-03-09

    Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling to mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were utilized to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and includes expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibit unique defect sink characteristics depending on their local structure. Furthermore, such interactions were found to be consistent across all doses investigated and had larger global implications including precipitation of Ni-Si clusters near different grain boundary types.

  14. Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

    DOE PAGES

    Field, Kevin G.; Yang, Ying; Busby, Jeremy T.; Allen, Todd R.

    2015-03-09

    Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling to mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were utilized to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed andmore » benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and includes expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibit unique defect sink characteristics depending on their local structure. Furthermore, such interactions were found to be consistent across all doses investigated and had larger global implications including precipitation of Ni-Si clusters near different grain boundary types.« less

  15. A method to calculate fission-fragment yields Y(Z,N) versus proton and neutron number in the Brownian shape-motion model. Application to calculations of U and Pu charge yields

    NASA Astrophysics Data System (ADS)

    Möller, Peter; Ichikawa, Takatoshi

    2015-12-01

    We propose a method to calculate the two-dimensional (2D) fission-fragment yield Y(Z,N) versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use the Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment Q2), neck d , left nascent fragment spheroidal deformation ɛ_{f1}, right nascent fragment deformation ɛ_{f2} and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalized potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of Z and N of the compound system and its shape, including the asymmetry of the shape. We outline here how to generalize the model from the "compound-system" model to a model where the emerging fragment proton and neutron numbers also enter, over and above the compound system composition.

  16. Comparison of whole-body phantom designs to estimate organ equivalent neutron doses for secondary cancer risk assessment in proton therapy

    NASA Astrophysics Data System (ADS)

    Moteabbed, Maryam; Geyer, Amy; Drenkhahn, Robert; Bolch, Wesley E.; Paganetti, Harald

    2012-01-01

    Secondary neutron fluence created during proton therapy can be a significant source of radiation exposure in organs distant from the treatment site, especially in pediatric patients. Various published studies have used computational phantoms to estimate neutron equivalent doses in proton therapy. In these simulations, whole-body patient representations were applied considering either generic whole-body phantoms or generic age- and gender-dependent phantoms. No studies to date have reported using patient-specific geometry information. The purpose of this study was to estimate the effects of patient-phantom matching when using computational pediatric phantoms. To achieve this goal, three sets of phantoms, including different ages and genders, were compared to the patients’ whole-body CT. These sets consisted of pediatric age-specific reference, age-adjusted reference and anatomically sculpted phantoms. The neutron equivalent dose for a subset of out-of-field organs was calculated using the GEANT4 Monte Carlo toolkit, where proton fields were used to irradiate the cranium and the spine of all phantoms and the CT-segmented patient models. The maximum neutron equivalent dose per treatment absorbed dose was calculated and found to be on the order of 0 to 5 mSv Gy-1. The relative dose difference between each phantom and their respective CT-segmented patient model for most organs showed a dependence on how close the phantom and patient heights were matched. The weight matching was found to have much smaller impact on the dose accuracy except for very heavy patients. Analysis of relative dose difference with respect to height difference suggested that phantom sculpting has a positive effect in terms of dose accuracy as long as the patient is close to the 50th percentile height and weight. Otherwise, the benefit of sculpting was masked by inherent uncertainties, i.e. variations in organ shapes, sizes and locations. Other sources of uncertainty included errors associated

  17. Effective dose of A-bomb radiation in Hiroshima and Nagasaki as assessed by chromosomal effectiveness of spectrum energy photons and neutrons.

    PubMed

    Sasaki, M S; Endo, S; Ejima, Y; Saito, I; Okamura, K; Oka, Y; Hoshi, M

    2006-07-01

    The effective dose of combined spectrum energy neutrons and high energy spectrum gamma-rays in A-bomb survivors in Hiroshima and Nagasaki has long been a matter of discussion. The reason is largely due to the paucity of biological data for high energy photons, particularly for those with an energy of tens of MeV. To circumvent this problem, a mathematical formalism was developed for the photon energy dependency of chromosomal effectiveness by reviewing a large number of data sets published in the literature on dicentric chromosome formation in human lymphocytes. The chromosomal effectiveness was expressed by a simple multiparametric function of photon energy, which made it possible to estimate the effective dose of spectrum energy photons and differential evaluation in the field of mixed neutron and gamma-ray exposure with an internal reference radiation. The effective dose of reactor-produced spectrum energy neutrons was insensitive to the fine structure of the energy distribution and was accessible by a generalized formula applicable to the A-bomb neutrons. Energy spectra of all sources of A-bomb gamma-rays at different tissue depths were simulated by a Monte Carlo calculation applied on an ICRU sphere. Using kerma-weighted chromosomal effectiveness of A-bomb spectrum energy photons, the effective dose of A-bomb neutrons was determined, where the relative biological effectiveness (RBE) of neutrons was expressed by a dose-dependent variable RBE, RBE(gamma, D (n)), against A-bomb gamma-rays as an internal reference radiation. When the newly estimated variable RBE(gamma, D (n)) was applied to the chromosome data of A-bomb survivors in Hiroshima and Nagasaki, the city difference was completely eliminated. The revised effective dose was about 35% larger in Hiroshima, 19% larger in Nagasaki and 26% larger for the combined cohort compared with that based on a constant RBE of 10. Since the differences are significantly large, the proposed effective dose might have an

  18. A method to calculate fission-fragment yields Y(Z,N) versus proton and neutron number in the Brownian shape-motion model

    SciTech Connect

    Moller, Peter; Ichikawa, Takatoshi

    2015-12-23

    In this study, we propose a method to calculate the two-dimensional (2D) fission-fragment yield Y(Z,N) versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use the Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment Q2), neck d, left nascent fragment spheroidal deformation ϵf1, right nascent fragment deformation ϵf2 and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalized potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of Z and N of the compound system and its shape, including the asymmetry of the shape. We outline here how to generalize the model from the “compound-system” model to a model where the emerging fragment proton and neutron numbers also enter, over and above the compound system composition.

  19. A method to calculate fission-fragment yields Y(Z,N) versus proton and neutron number in the Brownian shape-motion model

    DOE PAGES

    Moller, Peter; Ichikawa, Takatoshi

    2015-12-23

    In this study, we propose a method to calculate the two-dimensional (2D) fission-fragment yield Y(Z,N) versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use the Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment Q2), neck d, left nascent fragment spheroidal deformation ϵf1, right nascent fragment deformation ϵf2 and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalizedmore » potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of Z and N of the compound system and its shape, including the asymmetry of the shape. We outline here how to generalize the model from the “compound-system” model to a model where the emerging fragment proton and neutron numbers also enter, over and above the compound system composition.« less

  20. Neutron monitors and muon detectors for solar modulation studies: Interstellar flux, yield function, and assessment of critical parameters in count rate calculations

    NASA Astrophysics Data System (ADS)

    Maurin, D.; Cheminet, A.; Derome, L.; Ghelfi, A.; Hubert, G.

    2015-01-01

    Particles count rates at given Earth location and altitude result from the convolution of (i) the interstellar (IS) cosmic-ray fluxes outside the solar cavity, (ii) the time-dependent modulation of IS into Top-of-Atmosphere (TOA) fluxes, (iii) the rigidity cut-off (or geomagnetic transmission function) and grammage at the counter location, (iv) the atmosphere response to incoming TOA cosmic rays (shower development), and (v) the counter response to the various particles/energies in the shower. Count rates from neutron monitors or muon counters are therefore a proxy to solar activity. In this paper, we review all ingredients, discuss how their uncertainties impact count rate calculations, and how they translate into variation/uncertainties on the level of solar modulation ϕ (in the simple Force-Field approximation). The main uncertainty for neutron monitors is related to the yield function. However, many other effects have a significant impact, at the 5-10% level on ϕ values. We find no clear ranking of the dominant effects, as some depend on the station position and/or the weather and/or the season. An abacus to translate any variation of count rates (for neutron and μ detectors) to a variation of the solar modulation ϕ is provided.

  1. ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assesment Neutronics Analysis using the ATTILA Discrete Ordinates Code

    SciTech Connect

    Russell Feder and Mahmoud Z. Yousef

    2009-05-29

    Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of the ECH heating system were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture (ECH). The neutronics discrete-ordinates code ATTILA® and SEVERIAN® (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER “Brand Model” MCNP benchmark model. A biased quadrature set equivelant to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and ECH cases. The ECH or Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 μSv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 μSv/hr but fell below the limit to 90 μSv/hr 2-weeks later. The Large Aperture or ECH style shielding model exhibited higher and more persistent dose rates. After 1-day the dose rate was 230

  2. Comparative analysis of dose rates in bricks determined by neutron activation analysis, alpha counting and X-ray fluorescence analysis for the thermoluminescence fine grain dating method

    NASA Astrophysics Data System (ADS)

    Bártová, H.; Kučera, J.; Musílek, L.; Trojek, T.

    2014-11-01

    In order to evaluate the age from the equivalent dose and to obtain an optimized and efficient procedure for thermoluminescence (TL) dating, it is necessary to obtain the values of both the internal and the external dose rates from dated samples and from their environment. The measurements described and compared in this paper refer to bricks from historic buildings and a fine-grain dating method. The external doses are therefore negligible, if the samples are taken from a sufficient depth in the wall. However, both the alpha dose rate and the beta and gamma dose rates must be taken into account in the internal dose. The internal dose rate to fine-grain samples is caused by the concentrations of natural radionuclides 238U, 235U, 232Th and members of their decay chains, and by 40K concentrations. Various methods can be used for determining trace concentrations of these natural radionuclides and their contributions to the dose rate. The dose rate fraction from 238U and 232Th can be calculated, e.g., from the alpha count rate, or from the concentrations of 238U and 232Th, measured by neutron activation analysis (NAA). The dose rate fraction from 40K can be calculated from the concentration of potassium measured, e.g., by X-ray fluorescence analysis (XRF) or by NAA. Alpha counting and XRF are relatively simple and are accessible for an ordinary laboratory. NAA can be considered as a more accurate method, but it is more demanding regarding time and costs, since it needs a nuclear reactor as a neutron source. A comparison of these methods allows us to decide whether the time- and cost-saving simpler techniques introduce uncertainty that is still acceptable.

  3. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fallin, Brent; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Viera, David; Wilhelmy, Jerry

    2013-10-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He, or 3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers fission product yields relative to total number of fissions were determined at a high precision of 2-3% as well as absolute fission product yields at a lower precision of 5-6%. Results will be presented for a number of fission product isotopes at 1.38, 4.6 and 14.8 MeV as well as preliminary results at 9 MeV.

  4. Energy Dependence of Neutron-Induced Fission Product Yields of 235U, 238U and 239Pu Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Tornow, Werner; Tonchev, Anton; Vieira, Dave; Wilhelmy, Jerry; Arnold, Charles; Fowler, Malcolm; Stoyer, Mark

    2014-09-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements have been performed. The energy dependence of a number of cumulative fission products between 0.5 and 14.8 MeV have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of activation utilizing specially designed dual-fission chambers and γ-ray counting. The dual-fission chambers are back-to-back ionization chambers encasing a target with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the fission rate in the activation target with no reference to the fission cross-section, reducing uncertainties. γ-ray counting was performed on well-shield HPGe detectors over a period of 2 months per activation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 4.6 and 14.8 MeV.

  5. Differential neutron energy spectra measured on spacecraft low Earth orbit

    NASA Technical Reports Server (NTRS)

    Benton, E. V.; Frank, A. L.; Dudkin, E. V.; Potapov, Yu. V.; Akopova, A. B.; Melkumyan, L. V.

    1995-01-01

    Two methods for measuring neutrons in the range from thermal energies to dozens of MeV were used. In the first method, alpha-particles emitted from the (sup 6) Li(n.x)T reaction are detected with the help of plastic nuclear track detectors, yielding results on thermal and resonance neutrons. Also, fission foils are used to detect fast neutrons. In the second method, fast neutrons are recorded by nuclear photographic emulsions (NPE). The results of measurements on board various satellites are presented. The neutron flux density does not appear to correlate clearly with orbital parameters. Up to 50% of neutrons are due to albedo neutrons from the atmosphere while the fluxes inside the satellites are 15-20% higher than those on the outside. Estimates show that the neutron contribution to the total equivalent radiation dose reaches 20-30%.

  6. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

    DOE PAGES

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating themore » irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.« less

  7. Neutron source in the MCNPX shielding calculating for electron accelerator driven facility

    SciTech Connect

    Zhong, Z.; Gohar, Y.

    2012-07-01

    Argonne National Laboratory (ANL) of USA and Kharkov Inst. of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of an experimental neutron source facility. It is an accelerator driven system (ADS) utilizing a subcritical assembly driven by electron accelerator. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as a design tool due to its capability to transport electrons, photons, and neutrons at high energies. However the facility shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. A method, based on generating and utilizing neutron source file, was proposed and tested. This method reduces significantly the required computer resources and improves the statistics of the calculated neutron dose outside the shield boundary. However the statistical errors introduced by generating the neutron source were not directly represented in the results, questioning the validity of this methodology, because an insufficiently sampled neutron source can cause error on the calculated neutron dose. This paper presents a procedure for the validation of the generated neutron source file. The impact of neutron source statistic on the neutron dose is examined by calculating the neutron dose as a function of the number of electron particles used for generating the neutron source files. When the value of the calculated neutron dose converges, it means the neutron source has scored sufficient records and statistic does not have apparent impact on the calculated neutron dose. In this way, the validity of neutron source and the shield analyses could be verified. (authors)

  8. Mineralization- and remodeling-unrelated improvement of the post-yield properties of rat cortical bone by high doses of olpadronate.

    PubMed

    Capozza, R F; Mondelo, N; Reina, P S; Nocciolino, L; Meta, M; Roldan, E J; Ferretti, J L; Cointry, G R

    2013-06-01

    Some pharmacologic effects on bone modeling may not be evident in studies of remodeling skeletons. This study analyzes some effects of olpadronate on cortical bone modeling and post-yield properties in femurs diaphyses (virtually only-modeling bones) of young rats by mid-diaphyseal pQCT scans and bending tests. We studied 20/22 male/female animals traetad orally with olpadronate (45-90 mg/kg/d, 3 months) and 8/9 untreated controls. Both OPD doses enhanced diaphyseal cross-sectional moments of inertia (CSMI) with no change in cortical vBMD and elastic modulus. Yield stiffness and strength were mildly increased. Post-yield strength, deflection and energy absorption were strikingly enhanced. Ultimate strength was enhanced mainly because of effects on bone mass/geometry and post-yield properties. The large improvement of post-yield properties could be explained by improvements in bone geometry. Improvements in bone mass/geometry over weight-bearing needs suggest an enhanced modeling-related response to mechanical stimuli. Effects on tissue microstructural factors (not measured) could not be excluded. Results reveal novel olpadronate effects on bone strength and toughness unrelated to tissue mineralization and stiffness, even at high doses. Further studies could establish whether this could also occur in modeling-remodeling skeletons. If so, they could counteract the negative impact of anti-remodeling effects of bisphosphonates on bone strength.

  9. Corrigendum to "Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit"

    NASA Astrophysics Data System (ADS)

    El-Jaby, Samy

    2016-06-01

    A recent paper published in Life Sciences in Space Research (El-Jaby and Richardson, 2015) presented estimates of the secondary neutron ambient and effective dose equivalent rates, in air, from surface altitudes up to suborbital altitudes and low Earth orbit. These estimates were based on MCNPX (LANL, 2011) (Monte Carlo N-Particle eXtended) radiation transport simulations of galactic cosmic radiation passing through Earth's atmosphere. During a recent review of the input decks used for these simulations, a systematic error was discovered that is addressed here. After reassessment, the neutron ambient and effective dose equivalent rates estimated are found to be 10 to 15% different, though, the essence of the conclusions drawn remains unchanged.

  10. Corrigendum to "Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit".

    PubMed

    El-Jaby, Samy

    2016-06-01

    A recent paper published in Life Sciences in Space Research (El-Jaby and Richardson, 2015) presented estimates of the secondary neutron ambient and effective dose equivalent rates, in air, from surface altitudes up to suborbital altitudes and low Earth orbit. These estimates were based on MCNPX (LANL, 2011) (Monte Carlo N-Particle eXtended) radiation transport simulations of galactic cosmic radiation passing through Earth's atmosphere. During a recent review of the input decks used for these simulations, a systematic error was discovered that is addressed here. After reassessment, the neutron ambient and effective dose equivalent rates estimated are found to be 10 to 15% different, though, the essence of the conclusions drawn remains unchanged.

  11. Corrigendum to "Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit".

    PubMed

    El-Jaby, Samy

    2016-06-01

    A recent paper published in Life Sciences in Space Research (El-Jaby and Richardson, 2015) presented estimates of the secondary neutron ambient and effective dose equivalent rates, in air, from surface altitudes up to suborbital altitudes and low Earth orbit. These estimates were based on MCNPX (LANL, 2011) (Monte Carlo N-Particle eXtended) radiation transport simulations of galactic cosmic radiation passing through Earth's atmosphere. During a recent review of the input decks used for these simulations, a systematic error was discovered that is addressed here. After reassessment, the neutron ambient and effective dose equivalent rates estimated are found to be 10 to 15% different, though, the essence of the conclusions drawn remains unchanged. PMID:27345206

  12. [Radiolysis of liposomes, of solutions of ferrous sulphate and of albumin mixed by gamma-neutron radiation at different dose rates].

    PubMed

    Riabchenko, N I; Ul'ianenko, S E; Riabchenko, V I; Dzikovskaia, L A; Ivannik, V P; Sokolov, V A

    2005-01-01

    Solutions of the ferrous sulfate, of the albumin and of the suspension of liposomes were irradiated by mixed gamma-neutron radiation (fission spectrum neutrons, contribution of gamma-component to the absorbed dose up to 20%) at the pulse reactor BARS-6 with single-pulse (duration 100 micros) or continuous radiation (duration 60 min). It was shown, that after the pulse irradiation the concentration of the malonic dialdehyde in liposomes was in 3-4 times higher than after the continuous radiation at equal absorbed doses (p < 0.05). On the contrary, the irradiation of the liposomes suspension as well as of the solutions of the ferrous sulfate and of the albumin in a mode of single-pulse or of continuous mode did not reveal the statistically significant differences in the production of Fe3+ ions and of peroxides of the albumin for two mode of the radiation action. PMID:16304771

  13. Ultra-low-dose dual-source CT coronary angiography with high pitch: diagnostic yield of a volumetric planning scan and effects on dose reduction and imaging strategy

    PubMed Central

    Hamm, B; Huppertz, A; Lembcke, A

    2015-01-01

    Objective: To evaluate the role of an ultra-low-dose dual-source CT coronary angiography (CTCA) scan with high pitch for delimiting the range of the subsequent standard CTCA scan. Methods: 30 patients with an indication for CTCA were prospectively examined using a two-scan dual-source CTCA protocol (2.0 × 64.0 × 0.6 mm; pitch, 3.4; rotation time of 280 ms; 100 kV): Scan 1 was acquired with one-fifth of the tube current suggested by the automatic exposure control software [CareDose 4D™ (Siemens Healthcare, Erlangen, Germany) using 100 kV and 370 mAs as a reference] with the scan length from the tracheal bifurcation to the diaphragmatic border. Scan 2 was acquired with standard tube current extending with reduced scan length based on Scan 1. Nine central coronary artery segments were analysed qualitatively on both scans. Results: Scan 2 (105.1 ± 10.1 mm) was significantly shorter than Scan 1 (127.0 ± 8.7 mm). Image quality scores were significantly better for Scan 2. However, in 5 of 6 (83%) patients with stenotic coronary artery disease, a stenosis was already detected in Scan 1 and in 13 of 24 (54%) patients with non-stenotic coronary arteries, a stenosis was already excluded by Scan 1. Using Scan 2 as reference, the positive- and negative-predictive value of Scan 1 was 83% (5 of 6 patients) and 100% (13 of 13 patients), respectively. Conclusion: An ultra-low-dose CTCA planning scan enables a reliable scan length reduction of the following standard CTCA scan and allows for correct diagnosis in a substantial proportion of patients. Advances in knowledge: Further dose reductions are possible owing to a change in the individual patient's imaging strategy as a prior ultra-low-dose CTCA scan may already rule out the presence of a stenosis or may lead to a direct transferal to an invasive catheter procedure. PMID:25710210

  14. Measurement of neutron radiation exposure of commercial airline pilots using bubble detectors

    SciTech Connect

    Lewis, B.J.; Kosierb, R. . Dept. of Chemistry and Chemical Engineering); Cousins, T. . Space Systems and Technology Section); Hudson, D.F. ); Guery, G. )

    1994-06-01

    Neutron bubble detectors have been used over a 1-yr period by commercial airline pilots from Air Canada and Air France to measure the high-altitude neutron radiation exposure produced by galactic cosmic rays. The present work yielded measurements of the neutron flux of 1.0 to 4.6 n/cm[sup 2][center dot]s, and the neutron dose equivalent rates of 1.7 to 7.7 [mu]S[nu]/h. These measurements are in agreement with previous studies using high-altitude aircraft and conventional neutron instrumentation. The total dose equivalents for the Air Canada flights are also consistent with predictions of the CARI code. Considering that the neutron component contributes [approximately] 50% of the total dose equivalent, this study indicates that the annual dose for the air crew member would exceed the new recommendations of the International Commission on Radiological Protection (ICRP-60) for the general public.

  15. Reduced mutant yield at high doses in the Salmonella/activation assay: the cause is not always toxicity.

    PubMed

    McGregor, D; Prentice, R D; McConville, M; Lee, Y J; Caspary, W J

    1984-01-01

    In the Salmonella/activation assay developed by Ames et al [1973, 1975] toxicity is not measured, though it is recognized by the loss of a cloudy appearance on the plate. One approach to the measurement of toxicity is described here and uses a microscope-linked automated colony counter to estimate the number of microcolonies formed by histidine auxotrophs that stop growing after the depletion of histidine. This technique was used to evaluate the effect of toxicity on the revertant count for 16 mutagens, most of which were chosen because, from previous experience, their dose-response curves manifested a maximum at an intermediate dose tested. One of the sixteen, 2-nitrofluorene, was not toxic up to the maximum dose tested. The relationship between mutation and toxicity for the remaining fifteen allowed them to be grouped into two categories: (1) compounds that induced decreases in survival at the same dose at which the number of mutants decreased, and (2) compounds that induced toxicity, but survival was reduced at dose levels higher than those required to reduce the number of mutants. Possible explanations for this reduction of mutant counts occurring with little apparent concomitant increase in toxicity are examined. These results may be significant for attempts to estimate mutagenic potency and, to a lesser extent, construct mathematical models of the Ames test. PMID:6381041

  16. Modification of the University of Washington Neutron Radiotherapy Facility for optimization of neutron capture enhanced fast-neutron therapy.

    PubMed

    Nigg, D W; Wemple, C A; Risler, R; Hartwell, J K; Harker, Y D; Laramore, G E

    2000-02-01

    A modified neutron production target assembly has been developed to provide improved performance of the proton-cyclotron-based neutron radiotherapy facility at the University of Washington for applications involving neutron capture enhanced fast-neutron therapy. The new target produces a neutron beam that yields essentially the same fast-neutron physical depth-dose distribution as is produced by the current UW clinical system, but that also has an increased fraction of BNCT enhancement relative to the total therapeutic dose. The modified target is composed of a 5-millimeter layer of beryllium, followed by a 2.5-millimeter layer of tungsten, with a water-cooled copper backing. Measurements of the free-field neutron spectrum of the beam produced by the new target were performed using activation foils with a direct spectral unfolding technique. Water phantom measurements were performed using a tissue-equivalent ion chamber to characterize the fast-neutron depth-dose curve and sodium activation in soda-lime glass beads to characterize the thermal-neutron flux (and thus the expected neutron capture dose enhancement) as a function of depth. The results of the various measurements were quite consistent with expectations based on the design calculations for the modified target. The spectrum of the neutron beam produced by the new target features an enhanced low-energy flux component relative to the spectrum of the beam produced by the standard UW target. However, it has essentially the same high-energy neutron flux, with a reduced flux component in the mid-range of the energy spectrum. As a result, the measured physical depth-dose curve in a large water phantom has the same shape compared to the case of the standard UW clinical beam, but approximately twice the level of BNCT enhancement per unit background neutron dose at depths of clinical interest. In-vivo clinical testing of BNCT-enhanced fast-neutron therapy for canine lung tumors using the new beam was recently

  17. Modification of the University of Washington Neutron Radiotherapy Facility for optimization of neutron capture enhanced fast-neutron therapy

    SciTech Connect

    Nigg, David W.; Wemple, Charles A.; Risler, Ruedi; Hartwell, John K.; Harker, Yale D.; Laramore, George E.

    2000-02-01

    A modified neutron production target assembly has been developed to provide improved performance of the proton-cyclotron-based neutron radiotherapy facility at the University of Washington for applications involving neutron capture enhanced fast-neutron therapy. The new target produces a neutron beam that yields essentially the same fast-neutron physical depth-dose distribution as is produced by the current UW clinical system, but that also has an increased fraction of BNCT enhancement relative to the total therapeutic dose. The modified target is composed of a 5-millimeter layer of beryllium, followed by a 2.5-millimeter layer of tungsten, with a water-cooled copper backing. Measurements of the free-field neutron spectrum of the beam produced by the new target were performed using activation foils with a direct spectral unfolding technique. Water phantom measurements were performed using a tissue-equivalent ion chamber to characterize the fast-neutron depth-dose curve and sodium activation in soda-lime glass beads to characterize the thermal-neutron flux (and thus the expected neutron capture dose enhancement) as a function of depth. The results of the various measurements were quite consistent with expectations based on the design calculations for the modified target. The spectrum of the neutron beam produced by the new target features an enhanced low-energy flux component relative to the spectrum of the beam produced by the standard UW target. However, it has essentially the same high-energy neutron flux, with a reduced flux component in the mid-range of the energy spectrum. As a result, the measured physical depth-dose curve in a large water phantom has the same shape compared to the case of the standard UW clinical beam, but approximately twice the level of BNCT enhancement per unit background neutron dose at depths of clinical interest. In-vivo clinical testing of BNCT-enhanced fast-neutron therapy for canine lung tumors using the new beam was recently

  18. Exploratory study of fission product yields of neutron-induced fission of 235U , 238U , and 239Pu at 8.9 MeV

    NASA Astrophysics Data System (ADS)

    Bhatia, C.; Fallin, B. F.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E.; Bredeweg, T. A.; Fowler, M. M.; Moody, W.; Rundberg, R. S.; Rusev, G. Y.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2015-06-01

    Using dual-fission chambers each loaded with a thick (200 -400 -mg /c m2) actinide target of 235 ,238U or 239Pu and two thin (˜10 -100 -μ g /c m2) reference foils of the same actinide, the cumulative yields of fission products ranging from 92Sr to 147Nd have been measured at En= 8.9 MeV . The 2H(d ,n ) 3He reaction provided the quasimonoenergetic neutron beam. The experimental setup and methods used to determine the fission product yield (FPY) are described, and results for typically eight high-yield fission products are presented. Our FPYs for 235U(n ,f ) , 238U(n ,f ) , and 239Pu(n ,f ) at 8.9 MeV are compared with the existing data below 8 MeV from Glendenin et al. [Phys. Rev. C 24, 2600 (1981), 10.1103/PhysRevC.24.2600], Nagy et al. [Phys. Rev. C 17, 163 (1978), 10.1103/PhysRevC.17.163], Gindler et al. [Phys. Rev. C 27, 2058 (1983), 10.1103/PhysRevC.27.2058], and those of Mac Innes et al. [Nucl. Data Sheets 112, 3135 (2011), 10.1016/j.nds.2011.11.009] and Laurec et al. [Nucl. Data Sheets 111, 2965 (2010), 10.1016/j.nds.2010.11.004] at 14.5 and 14.7 MeV, respectively. This comparison indicates a negative slope for the energy dependence of most fission product yields obtained from 235U and 239Pu , whereas for 238U the slope issue remains unsettled.

  19. Measurement of charged particle yields from therapeutic beams in view of the design of an innovative hadrontherapy dose monitor

    NASA Astrophysics Data System (ADS)

    Battistoni, G.; Bellini, F.; Bini, F.; Collamati, F.; Collini, F.; De Lucia, E.; Durante, M.; Faccini, R.; Ferroni, F.; Frallicciardi, P. M.; La Tessa, C.; Marafini, M.; Mattei, I.; Miraglia, F.; Morganti, S.; Ortega, P. G.; Patera, V.; Piersanti, L.; Pinci, D.; Russomando, A.; Sarti, A.; Schuy, C.; Sciubba, A.; Senzacqua, M.; Solfaroli Camillocci, E.; Vanstalle, M.; Voena, C.

    2015-02-01

    Particle Therapy (PT) is an emerging technique, which makes use of charged particles to efficiently cure different kinds of solid tumors. The high precision in the hadrons dose deposition requires an accurate monitoring to prevent the risk of under-dosage of the cancer region or of over-dosage of healthy tissues. Monitoring techniques are currently being developed and are based on the detection of particles produced by the beam interaction into the target, in particular: charged particles, result of target and/or projectile fragmentation, prompt photons coming from nucleus de-excitation and back-to-back γ s, produced in the positron annihilation from β + emitters created in the beam interaction with the target. It has been showed that the hadron beam dose release peak can be spatially correlated with the emission pattern of these secondary particles. Here we report about secondary particles production (charged fragments and prompt γ s) performed at different beam and energies that have a particular relevance for PT applications: 12C beam of 80 MeV/u at LNS, 12C beam 220 MeV/u at GSI, and 12C, 4He, 16O beams with energy in the 50-300 MeV/u range at HIT. Finally, a project for a multimodal dose-monitor device exploiting the prompt photons and charged particles emission will be presented.

  20. Recovery capacity of glial progenitors after in vivo fission-neutron or X irradiation: age dependence, fractionation and low-dose-rate irradiations.

    PubMed

    Philippo, H; Winter, E A M; van der Kogel, A J; Huiskamp, R

    2005-06-01

    Previous experiments on the radiosensitivity of O-2A glial progenitors determined for single-dose fission-neutron and X irradiation showed log-linear survival curves, suggesting a lack of accumulation of recovery of sublethal damage. In the present study, we addressed this question and further characterized the radiobiological properties of these glial stem cells by investigating the recovery capacity of glial stem cells using either fractionated or protracted whole-body irradiation. Irradiations were performed on newborn, 2-week-old or 12-week-old rats. Fractionated irradiations (four fractions) were performed with 24-h intervals, followed by cell isolations 16- 24 h after the last irradiation. Single-dose irradiations were followed by cell isolation 16-24 h after irradiation or delayed cell isolation (4 days after irradiation) of the O-2A progenitor cells from either spinal cord (newborns) or optic nerve (2- and 12-week-old rats). Results for neonatal progenitor cell survival show effect ratios for both fractionated fission-neutron and X irradiation of the order of 1.8 when compared with single-dose irradiation. A similar ratio was found after single-dose irradiation combined with delayed plating. Comparable results were observed for juvenile and adult optic nerve progenitors, with effect ratios of the order of 1.2. The present investigation clearly shows that fractionated irradiation regimens using X rays or fission neutrons and CNS tissue from rats of various ages results in an increase in O-2A progenitor cell survival while repair is virtually absent. This recovery of the progenitor pool after irradiation can be observed at all ages but is greatest in the neonatal spinal cord and can probably be attributed to repopulation. PMID:15913395

  1. Low-temperature low-dose neutron irradiation effects on Brush Wellman S65-C and Kawechi Berylco P0 beryllium

    SciTech Connect

    Snead, L.L.

    1998-09-01

    The mechanical property results for two high quality beryllium materials subjected to low temperature, low dose neutron irradiation in water moderated reactors are presented. Materials chosen were the S65-C ITER candidate material produced by Brush Wellman, and Kawecki Berylco Industries P0 beryllium. Both materials were processed by vacuum hot pressing. Mini sheet tensile and thermal diffusivity specimens were irradiated in the temperature range of {approximately}100--275 C from a fast (E > 0.1 MeV) neutron dose of 0.05 to 1.0 {times} 10{sup 25} n/m{sup 2} in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory and the High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory. As expected from earlier work on beryllium, both materials underwent significant embrittlement with corresponding reduction in ductility and increased strength. Both thermal diffusivity and volumetric expansion were measured and found to be negligible in this temperature and fluence range. Of significance from this work is that while both materials rapidly embrittle at these ITER relevant irradiation conditions, some ductility (>1--2%) remains, which contrasts with a body of earlier work including recent work on the Brush-Wellman S65-C material irradiated to slightly higher neutron fluence.

  2. Thick target D-T neutron yield measurements using metal occluders of scandium, titanium, yttrium, zirconium, gadolinium, erbium, hafnium, and tantalum at energies from 25 to 200 keV

    SciTech Connect

    Malbrough, D.J.; Molloy, J.T. Jr.; Becker, R.H.

    1990-11-19

    Deuterium-Tritium (D-T) neutron yields from thick films of scandium, titanium, yttrium, zirconium, gadolinium, erbium, hafnium, and tantalum were measured by the associated particle technique using the 200-keV accelerator at the Pinellas Plant. The neutron yields were measured for all targets at energies from 25 to 200 keV in 5-keV steps with an average uncertainty of {plus_minus}6.8%. Tabulated results are presented with yield versus energy curves for each target. Yield curves for D-D neutrons from earlier measurements are also presented with the D-T neutron yield curves. Good fits to the data were found for both D-D and D-T with theoretical calculations that were adjusted by smooth functions of the form: A{sub 0} + A{sub 1}E + A{sub 2}E{sup 2}. The results of the fits strongly suggest that disagreement between measurement and theory is due mainly to inaccuracies in currently available stopping power data. Comparisons with earlier theoretical calculations for titanium and erbium are also presented. 27 refs., 10 figs., 4 tabs.

  3. Fluence-to-dose conversion coefficients for neutrons and protons calculated using the PHITS code and ICRP/ICRU adult reference computational phantoms.

    PubMed

    Sato, Tatsuhiko; Endo, Akira; Zankl, Maria; Petoussi-Henss, Nina; Niita, Koji

    2009-04-01

    The fluence to organ-dose and effective-dose conversion coefficients for neutrons and protons with energies up to 100 GeV was calculated using the PHITS code coupled to male and female adult reference computational phantoms, which are to be released as a common ICRP/ICRU publication. For the calculation, the radiation and tissue weighting factors, w(R) and w(T), respectively, as revised in ICRP Publication 103 were employed. The conversion coefficients for effective dose equivalents derived using the radiation quality factors of both Q(L) and Q(y) relationships were also estimated, utilizing the functions for calculating the probability densities of the absorbed dose in terms of LET (L) and lineal energy (y), respectively, implemented in PHITS. By comparing these data with the corresponding data for the effective dose, we found that the numerical compatibilities of the revised w(R) with the Q(L) and Q(y) relationships are fairly established. The calculated data of these dose conversion coefficients are indispensable for constructing the radiation protection systems based on the new recommendations given in ICRP103 for aircrews and astronauts, as well as for workers in accelerators and nuclear facilities.

  4. Affinity and dose of TCR engagement yield proportional enhancer and gene activity in CD4+ T cells

    PubMed Central

    Allison, Karmel A; Sajti, Eniko; Collier, Jana G; Gosselin, David; Troutman, Ty Dale; Stone, Erica L; Hedrick, Stephen M; Glass, Christopher K

    2016-01-01

    Affinity and dose of T cell receptor (TCR) interaction with antigens govern the magnitude of CD4+ T cell responses, but questions remain regarding the quantitative translation of TCR engagement into downstream signals. We find that while the response of mouse CD4+ T cells to antigenic stimulation is bimodal, activated cells exhibit analog responses proportional to signal strength. Gene expression output reflects TCR signal strength, providing a signature of T cell activation. Expression changes rely on a pre-established enhancer landscape and quantitative acetylation at AP-1 binding sites. Finally, we show that graded expression of activation genes depends on ERK pathway activation, suggesting that an ERK-AP-1 axis plays an important role in translating TCR signal strength into proportional activation of enhancers and genes essential for T cell function. DOI: http://dx.doi.org/10.7554/eLife.10134.001 PMID:27376549

  5. SU-E-T-403: Measurement of the Neutron Ambient Dose Equivalent From the TrueBeam Linac Head and Varian 2100 Clinac

    SciTech Connect

    Harvey, M; Pollard, J; Wen, Z; Gao, S

    2014-06-01

    Purpose: High-energy x-ray therapy produces an undesirable source of stray neutron dose to healthy tissues, and thus, poses a risk for second cancer induction years after the primary treatment. Hence, the purpose of this study was to measure the neutron ambient dose equivalent, H*(10), produced from the TrueBeam and Varian 2100 linac heads, respectively. Of particular note is that there is no measured data available in the literature on H*(10) production from the TrueBeam treatment head. Methods: Both linacs were operated in flattening filter mode using a 15 MV x-ray beam on TrueBeam and an 18 MV x-ray beam for the Varian 2100 Clinac with the jaws and multileaf collimators in the fully closed position. A dose delivery rate of 600 MU/min was delivered on the TrueBeam and the Varian 2100 Clinac, respectively and the H*(10) rate was measured in triplicate using the WENDI-2 detector located at multiple positions including isocenter and longitudinal (gun-target) to the isocenter. Results: For each measurement, the H*(10) rate was relatively constant with increasing distance away from the isocenter with standard deviations on the order of a tenth of a mSv/h or less for the given beam energy. In general, fluctuations in the longitudinal H*(10) rate between the anterior-posterior couch directions were approximately a percent for both beam energies. Conclusion: Our preliminary results suggest an H*(10) rate of about 30 mSv/h (40 mSv/h) or less for TrueBeam (Varian Clinac 2100) for all measurements considered in this study indicating a relatively low contribution of produced secondary neutrons to the primary therapeutic beam.

  6. The Ratio R{sub dp} of the quasielastic nd {yields} p(nn) to the elastic np {yields} pn charge-exchange-process yields at the proton emitting angle {theta}{sub p,lab} = 0 deg. over 0.55-2.0 GeV neutron-beam energy region. Comparison of the results with the model-dependent calculations

    SciTech Connect

    Sharov, V. I. Morozov, A. A.; Shindin, R. A.; Chernykh, E. V.; Nomofilov, A. A.; Strunov, L. N.

    2009-06-15

    Our new experimental results (see, e.g., Preprint JINR no. E1-2008-61 (Dubna, 2008)) on ratio R{sub dp} of the quasielastic charge-exchange yield at the proton emitting angle {theta}{sub p,lab} = 0 deg. for the nd {yields} p(nn) reaction to the elastic np {yields} pn charge-exchange yield were presented. The measurements were carried out at the Nuclotron of the Veksler and Baldin Laboratory of High Energies of the JINR (Dubna) at the neutron-beam kinetic energies of 0.55, 0.8, 1.0, 1.2, 1.4, 1.8, and 2.0 GeV. In this paper the comparison of the experimental R{sub dp} data with the obtained R{sub dp} calculations within the impulse approximation by using the invariant-amplitude sets from the GW/VPI phase-shift analysis is made. The R{sub dp} values calculated using the set of invariant amplitude data for the elastic np {yields} pn charge exchange at {theta}{sub p,CM} = 0 deg., agree with the experimental data. This confirmed the nd {yields} p(nn) process yield at {theta}{sub p,CM} = 0 deg. is caused by the contribution of the spin-dependent part of the elastic np {yields} pn charge-exchange reaction. Thus, it has been shown that the obtained experimental R{sub dp} results can be used for the Delta-Sigma experimental program to reduce the total ambiguity in the extraction of the amplitude real parts.

  7. 32 CFR 218.4 - Dose estimate reporting standards.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ..., and neutron doses, when applicable. In determining the veteran's dose, initial neutron, initial gamma..., doses will be reported as gamma dose, neutron dose, and internal dose. To the extent to which the... of a neutron or internal exposure? What is the reconstruction? Upon request, the participant or...

  8. 32 CFR 218.4 - Dose estimate reporting standards.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ..., and neutron doses, when applicable. In determining the veteran's dose, initial neutron, initial gamma..., doses will be reported as gamma dose, neutron dose, and internal dose. To the extent to which the... of a neutron or internal exposure? What is the reconstruction? Upon request, the participant or...

  9. 32 CFR 218.4 - Dose estimate reporting standards.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ..., and neutron doses, when applicable. In determining the veteran's dose, initial neutron, initial gamma..., doses will be reported as gamma dose, neutron dose, and internal dose. To the extent to which the... of a neutron or internal exposure? What is the reconstruction? Upon request, the participant or...

  10. 32 CFR 218.4 - Dose estimate reporting standards.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ..., and neutron doses, when applicable. In determining the veteran's dose, initial neutron, initial gamma..., doses will be reported as gamma dose, neutron dose, and internal dose. To the extent to which the... of a neutron or internal exposure? What is the reconstruction? Upon request, the participant or...

  11. PERSONNEL NEUTRON DOSIMETER

    DOEpatents

    Fitzgerald, J.J.; Detwiler, C.G. Jr.

    1960-05-24

    A description is given of a personnel neutron dosimeter capable of indicating the complete spectrum of the neutron dose received as well as the dose for each neutron energy range therein. The device consists of three sets of indium foils supported in an aluminum case. The first set consists of three foils of indium, the second set consists of a similar set of indium foils sandwiched between layers of cadmium, whereas the third set is similar to the second set but is sandwiched between layers of polyethylene. By analysis of all the foils the neutron spectrum and the total dose from neutrons of all energy levels can be ascertained.

  12. Silver fluorescent x-ray yield and its influence on the dose rate constant for nine low-energy brachytherapy source models

    SciTech Connect

    Nath, Ravinder; Chen, Zhe Jay

    2007-10-15

    The physical characteristics of the photons emitted by a low-energy brachytherapy source are strongly dependent on the source's construction. Aside from absorption and scattering caused by the internal structures and the source encapsulation, the photoelectric interactions occurred in certain type of source-construction materials can generate additional energetic characteristic x rays with energies different from those emitted by the bare radionuclide. As a result, the same radionuclide encapsulated in different source models can result in dose rate constants and other dosimetric parameters that are strikingly different from each other. The aim of this work was to perform a systematic study on the yield of silver fluorescent x rays produced in nine {sup 125}I sources that are known to contain silver and its impact on the dose-rate constant. Using a high-resolution germanium spectrometer, the relative {sup 125}I spectra emitted by the nine sources on its bisector were measured and found to be similar to each other (the maximum variation in the {sup 125}I-K{sub {beta}} relative intensity was less than 4%). On the other hand, the measured silver fluorescent x-ray spectra exhibited much greater variations from model to model; the maximum change in the measured Ag-K{sub {alpha}} relative intensity was over 95%. This larger variation in the measured silver fluorescent x-ray yield was caused by (1) the different amount of silver that was directly exposed to the {sup 125}I radionuclide in different source models and (2) the stronger influence of the source's internal geometry on the silver fluorescent x rays. Because the addition of silver fluorescent x rays can significantly alter the photon characteristics emitted by the radioactive sources, a precise knowledge on the silver fluorescent x-ray yield is needed in theoretical calculations of the sources' intrinsic dosimetric properties. This study concludes that the differences in silver fluorescent yield are the primary

  13. Neutron Tube Design Study for Boron Neutron Capture TherapyApplication

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1998-01-04

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  14. Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

    NASA Astrophysics Data System (ADS)

    Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Bélier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric; Alvarez-Pol, Héctor; Audouin, Laurent; Aumann, Thomas; Ayyad, Yassid; Benlliure, Jose; Casarejos, Enrique; Cortina Gil, Dolores; Caamaño, Manuel; Farget, Fanny; Fernández Domínguez, Beatriz; Heinz, Andreas; Jurado, Beatriz; Kelić-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Paradela, Carlos; Pietri, Stéphane; Ramos, Diego; Rodríguez-Sànchez, Jose-Luis; Rodríguez-Tajes, Carme; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Tassan-Got, Laurent; Vargas, Jossitt; Voss, Bernd; Weick, Helmut

    2015-12-01

    A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism.

  15. Proton recoil scintillator neutron rem meter

    DOEpatents

    Olsher, Richard H.; Seagraves, David T.

    2003-01-01

    A neutron rem meter utilizing proton recoil and thermal neutron scintillators to provide neutron detection and dose measurement. In using both fast scintillators and a thermal neutron scintillator the meter provides a wide range of sensitivity, uniform directional response, and uniform dose response. The scintillators output light to a photomultiplier tube that produces an electrical signal to an external neutron counter.

  16. Differential neutron energy spectra measured on spacecraft in low Earth orbit.

    PubMed

    Dudkin, V E; Potapov YuV; Akopova, A B; Melkumyan, L V; Benton, E V; Frank, A L

    1990-01-01

    Two methods for measuring neutrons in the range from thermal energies to dozens of MeV were used. In the first method, alpha-particles emitted from the 6Li(n,alpha)T reaction are detected with the help of plastic nuclear track detectors, yielding results on thermal and resonance neutrons. Also, fission foils are used to detect fast neutrons. In the second method, fast neutrons are recorded by nuclear photographic emulsions (NPE). The results of measurements on board various satellites are presented. The neutron flux density does not appear to correlate clearly with orbital parameters. Up to 50% of neutrons are due to albedo neutrons from the atmosphere while the fluxes inside the satellites are 15-20% higher than those on the outside. Estimates show that the neutron contribution to the total equivalent radiation dose reaches 20-30%. PMID:11537519

  17. Differential neutron energy spectra measured on spacecraft in low Earth orbit

    NASA Technical Reports Server (NTRS)

    Dudkin, V. E.; Akopova, A. B.; Melkumyan, L. V.; Benton, E. V.; Frank, A. L.

    1990-01-01

    Two methods for measuring neutrons in the range from thermal energies to dozens of MeV were used. In the first method, alpha-particles emitted from the 6Li(n,alpha)T reaction are detected with the help of plastic nuclear track detectors, yielding results on thermal and resonance neutrons. Also, fission foils are used to detect fast neutrons. In the second method, fast neutrons are recorded by nuclear photographic emulsions (NPE). The results of measurements on board various satellites are presented. The neutron flux density does not appear to correlate clearly with orbital parameters. Up to 50% of neutrons are due to albedo neutrons from the atmosphere while the fluxes inside the satellites are 15-20% higher than those on the outside. Estimates show that the neutron contribution to the total equivalent radiation dose reaches 20-30%.

  18. Neutron and X-ray diffraction analysis of the effect of irradiation dose and temperature on microstructure of irradiated HT-9 steel

    NASA Astrophysics Data System (ADS)

    Mosbrucker, P. L.; Brown, D. W.; Anderoglu, O.; Balogh, L.; Maloy, S. A.; Sisneros, T. A.; Almer, J.; Tulk, E. F.; Morgenroth, W.; Dippel, A. C.

    2013-11-01

    Material harvested from several positions within a nuclear fuel duct (the ACO-3 duct) used in a 6-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF) was examined using neutron and high-energy X-ray diffraction. Samples with a wide range of irradiation dose and irradiation temperature history, reaching doses of up to 147 dpa and temperatures of up to 777 K, were examined. The response of various microstructural characteristics such as the weight fraction of M23C6 carbides, the dislocation density and character, and the crystallographic texture were determined using whole profile analysis of the diffraction data and related to the macroscopic mechanical behavior. For instance, the dislocation density was observed to be intimately linked with observed flow strength of the irradiated materials, following the Taylor law. In general, at the high doses studied in this work, the irradiation temperature is the predominant controlling factor of the dislocation density and, thus, the flow strength of the irradiated material. The results, representing some of the first diffraction work done on samples exposed to such a high received dose, demonstrate how non-destructive and stand-off diffraction techniques can be used to characterize irradiation induced microstructure and at least estimate mechanical properties in irradiated materials without exposing workers to radiation hazards.

  19. Microtron MT 25 as a source of neutrons

    SciTech Connect

    Kralik, M.; Solc, J.; Chvatil, D.; Krist, P.; Turek, K.; Granja, C.

    2012-08-15

    The objective was to describe Microtron MT25 as a source of neutrons generated by bremsstrahlung induced photonuclear reactions in U and Pb targets. Bremsstrahlung photons were produced by electrons accelerated at energy 21.6 MeV. Spectral fluence of the generated neutrons was calculated with MCNPX code and then experimentally determined at two positions by means of a Bonner spheres spectrometer in which the detector of thermal neutrons was replaced by activation Mn tablets or track detectors CR-39 with a {sup 10}B radiator. The measured neutron spectral fluence and the calculated anisotropy served for the estimation of neutron yield from the targets and for the determination of ambient dose equivalent rate at the place of measurement. Microtron MT25 is intended as one of the sources for testing neutron sensitive devices which will be sent into the space.

  20. NEUTRON SOURCES

    DOEpatents

    Richmond, J.L.; Wells, C.E.

    1963-01-15

    A neutron source is obtained without employing any separate beryllia receptacle, as was formerly required. The new method is safer and faster, and affords a source with both improved yield and symmetry of neutron emission. A Be container is used to hold and react with Pu. This container has a thin isolating layer that does not obstruct the desired Pu--Be reaction and obviates procedures previously employed to disassemble and remove a beryllia receptacle. (AEC)

  1. [Results of measuring neutrons doses and energy spectra inside Russian segment of the International Space Station in experiment "Matryoshka-R" using bubble detectors during the ISS-24-34 missions].

    PubMed

    Khulapko, S V; Liagushin, V I; Arkhangel'skiĭ, V V; Shurshakov, V A; Smith, M; Ing, H; Machrafi, R; Nikolaev, I V

    2014-01-01

    The paper presents the results of calculating the equivalent dose from and energy spectrum of neutrons in the right-hand crewquarters in module Zvezda of the ISS Russian segment. Dose measurements were made in the period between July, 2010 and November, 2012 (ISS Missions 24-34) by research equipment including the bubble dosimeter as part of experiment "Matryoshka-R". Neutron energy spectra in the crewquarters are in good agreement with what has been calculated for the ISS USOS and, earlier, for the MIR orbital station. The neutron dose rate has been found to amount to 196 +/- 23 microSv/d on Zvezda panel-443 (crewquarters) and 179 +/- 16 microSv/d on the "Shielding shutter" surface in the crewquarters.

  2. A high yield neutron target

    NASA Technical Reports Server (NTRS)

    Alger, D. L.; Steinberg, R.; Weisenbach, P.

    1974-01-01

    Target, in cylinder form, rotates rapidly in front of beam. Titanium tritide film is much thicker than range of accelerated deutron. Sputtering electrode permits full use of thick film. Stream of high-velocity coolant provides efficient transfer of heat from target.

  3. Using the TREAT reactor in support of boron neutron capture therapy (BNCT) experiments: A feasibility analysis

    SciTech Connect

    Grasseschi, G.L.; Schaefer, R.W.

    1996-03-01

    The technical feasibility of using the TREAT reactor facility for boron neutron capture therapy (BNCT) research was assessed. Using one-dimensional neutronics calculations, it was shown that the TREAT core neutron spectrum can be filtered to reduce the undesired radiation (contamination) dose per desired neutron more effectively than can the core spectra from two prominent candidate reactors. Using two-dimensional calculations, it was demonstrated that a non-optimized filter replacing the TREAT thermal column can yield a fluence of desired-energy neutrons more than twice as large as the fluence believed to be required and, at the same time, have a contamination dose per desired neutron almost as low as that from any other candidate facility. The time, effort and cost required to adapt TREAT for a mission supporting BNCT research would be modest.

  4. Fluence-to-absorbed-dose conversion coefficients for neutron beams from 0.001 eV to 100 GeV calculated for a set of pregnant female and fetus models

    NASA Astrophysics Data System (ADS)

    Taranenko, Valery; Xu, X. George

    2008-03-01

    Protection of fetuses against external neutron exposure is an important task. This paper reports a set of absorbed dose conversion coefficients for fetal and maternal organs for external neutron beams using the RPI-P pregnant female models and the MCNPX code. The newly developed pregnant female models represent an adult female with a fetus including its brain and skeleton at the end of each trimester. The organ masses were adjusted to match the reference values within 1%. For the 3 mm cubic voxel size, the models consist of 10-15 million voxels for 35 organs. External monoenergetic neutron beams of six standard configurations (AP, PA, LLAT, RLAT, ROT and ISO) and source energies 0.001 eV-100 GeV were considered. The results are compared with previous data that are based on simplified anatomical models. The differences in dose depend on source geometry, energy and gestation periods: from 20% up to 140% for the whole fetus, and up to 100% for the fetal brain. Anatomical differences are primarily responsible for the discrepancies in the organ doses. For the first time, the dependence of mother organ doses upon anatomical changes during pregnancy was studied. A maximum of 220% increase in dose was observed for the placenta in the nine months model compared to three months, whereas dose to the pancreas, small and large intestines decreases by 60% for the AP source for the same models. Tabulated dose conversion coefficients for the fetus and 27 maternal organs are provided.

  5. Evaluation of all dose components in the LVR-15 reactor epithermal neutron beam using Fricke gel dosimeter layers.

    PubMed

    Bartesaghi, G; Burian, J; Gambarini, G; Marek, M; Negri, A; Viererbl, L

    2009-07-01

    Fricke gel dosimeters in the form of layers are suitable to reconstruct bidimensional distributions of the absorbed dose; in accordance with their chemical composition and applying suitably developed algorithms, they can provide dose images of the different radiation components in a BNCT field. After the description of the applied method, this work presents the results obtained at the epithermal column of the BNCT facility at the NRI in Rez (CZ). The measured dose distributions are shown in comparison with data taken by means of other dosimeters thermoluminescence dosimeters (TLDs) and with calculations carried out with the Monte Carlo code MCNP5. The agreement with the results obtained by means of the different techniques is satisfying.

  6. Impact of intra-arterial administration of boron compounds on dose-volume histograms in boron neutron capture therapy for recurrent head-and-neck tumors

    SciTech Connect

    Suzuki, Minoru . E-mail: msuzuki@rri.kyoto-u.ac.jp; Sakurai, Yoshinori; Nagata, Kenji; Kinashi, Yuko; Masunaga, Shinichiro; Ono, Koji; Maruhashi, Akira; Kato, Ituro; Fuwa, Nobukazu; Hiratsuka, Junichi; Imahori, Yoshio

    2006-12-01

    Purpose: To analyze the dose-volume histogram (DVH) of head-and-neck tumors treated with boron neutron capture therapy (BNCT) and to determine the advantage of the intra-arterial (IA) route over the intravenous (IV) route as a drug delivery system for BNCT. Methods and Materials: Fifteen BNCTs for 12 patients with recurrent head-and-neck tumors were included in the present study. Eight irradiations were done after IV administration of boronophenylalanine and seven after IA administration. The maximal, mean, and minimal doses given to the gross tumor volume were assessed using a BNCT planning system. Results: The results are reported as median values with the interquartile range. In the IA group, the maximal, mean, and minimal dose given to the gross tumor volume was 68.7 Gy-Eq (range, 38.8-79.9), 45.0 Gy-Eq (range, 25.1-51.0), and 13.8 Gy-Eq (range, 4.8-25.3), respectively. In the IV group, the maximal, mean, and minimal dose given to the gross tumor volume was 24.2 Gy-Eq (range, 21.5-29.9), 16.4 Gy-Eq (range, 14.5-20.2), and 7.8 Gy-Eq (range, 6.8-9.5), respectively. Within 1-3 months after BNCT, the responses were assessed. Of the 6 patients in the IV group, 2 had a partial response, 3 no change, and 1 had progressive disease. Of 4 patients in the IA group, 1 achieved a complete response and 3 a partial response. Conclusion: Intra-arterial administration of boronophenylalanine is a promising drug delivery system for head-and-neck BNCT.

  7. Accelerator-Based Biological Irradiation Facility Simulating Neutron Exposure from an Improvised Nuclear Device

    PubMed Central

    Xu, Yanping; Randers-Pehrson, Gerhard; Turner, Helen C.; Marino, Stephen A.; Geard, Charles R.; Brenner, David J.; Garty, Guy

    2015-01-01

    We describe here an accelerator-based neutron irradiation facility, intended to expose blood or small animals to neutron fields mimicking those from an improvised nuclear device at relevant distances from the epicenter. Neutrons are generated by a mixed proton/deuteron beam on a thick beryllium target, generating a broad spectrum of neutron energies that match those estimated for the Hiroshima bomb at 1.5 km from ground zero. This spectrum, dominated by neutron energies between 0.2 and 9 MeV, is significantly different from the standard reactor fission spectrum, as the initial bomb spectrum changes when the neutrons are transported through air. The neutron and gamma dose rates were measured using a custom tissue-equivalent gas ionization chamber and a compensated Geiger-Mueller dosimeter, respectively. Neutron spectra were evaluated by unfolding measurements using a proton-recoil proportional counter and a liquid scintillator detector. As an illustration of the potential use of this facility we present micronucleus yields in single divided, cytokinesis-blocked human peripheral lymphocytes up to 1.5 Gy demonstrating 3- to 5-fold enhancement over equivalent X-ray doses. This facility is currently in routine use, irradiating both mice and human blood samples for evaluation of neutron-specific biodosimetry assays. Future studies will focus on dose reconstruction in realistic mixed neutron/photon fields. PMID:26414507

  8. Accelerator-Based Biological Irradiation Facility Simulating Neutron Exposure from an Improvised Nuclear Device.

    PubMed

    Xu, Yanping; Randers-Pehrson, Gerhard; Turner, Helen C; Marino, Stephen A; Geard, Charles R; Brenner, David J; Garty, Guy

    2015-10-01

    We describe here an accelerator-based neutron irradiation facility, intended to expose blood or small animals to neutron fields mimicking those from an improvised nuclear device at relevant distances from the epicenter. Neutrons are generated by a mixed proton/deuteron beam on a thick beryllium target, generating a broad spectrum of neutron energies that match those estimated for the Hiroshima bomb at 1.5 km from ground zero. This spectrum, dominated by neutron energies between 0.2 and 9 MeV, is significantly different from the standard reactor fission spectrum, as the initial bomb spectrum changes when the neutrons are transported through air. The neutron and gamma dose rates were measured using a custom tissue-equivalent gas ionization chamber and a compensated Geiger-Mueller dosimeter, respectively. Neutron spectra were evaluated by unfolding measurements using a proton-recoil proportional counter and a liquid scintillator detector. As an illustration of the potential use of this facility we present micronucleus yields in single divided, cytokinesis-blocked human peripheral lymphocytes up to 1.5 Gy demonstrating 3- to 5-fold enhancement over equivalent X-ray doses. This facility is currently in routine use, irradiating both mice and human blood samples for evaluation of neutron-specific biodosimetry assays. Future studies will focus on dose reconstruction in realistic mixed neutron/photon fields. PMID:26414507

  9. Accelerator-Based Biological Irradiation Facility Simulating Neutron Exposure from an Improvised Nuclear Device.

    PubMed

    Xu, Yanping; Randers-Pehrson, Gerhard; Turner, Helen C; Marino, Stephen A; Geard, Charles R; Brenner, David J; Garty, Guy

    2015-10-01

    We describe here an accelerator-based neutron irradiation facility, intended to expose blood or small animals to neutron fields mimicking those from an improvised nuclear device at relevant distances from the epicenter. Neutrons are generated by a mixed proton/deuteron beam on a thick beryllium target, generating a broad spectrum of neutron energies that match those estimated for the Hiroshima bomb at 1.5 km from ground zero. This spectrum, dominated by neutron energies between 0.2 and 9 MeV, is significantly different from the standard reactor fission spectrum, as the initial bomb spectrum changes when the neutrons are transported through air. The neutron and gamma dose rates were measured using a custom tissue-equivalent gas ionization chamber and a compensated Geiger-Mueller dosimeter, respectively. Neutron spectra were evaluated by unfolding measurements using a proton-recoil proportional counter and a liquid scintillator detector. As an illustration of the potential use of this facility we present micronucleus yields in single divided, cytokinesis-blocked human peripheral lymphocytes up to 1.5 Gy demonstrating 3- to 5-fold enhancement over equivalent X-ray doses. This facility is currently in routine use, irradiating both mice and human blood samples for evaluation of neutron-specific biodosimetry assays. Future studies will focus on dose reconstruction in realistic mixed neutron/photon fields.

  10. The Dose-Effect Relationship Between the Seeding Quantity of Human Marrow Mesenchymal Stem Cells and In Vivo Tissue-Engineered Bone Yield.

    PubMed

    Wu, Huanhuan; Kang, Ning; Wang, Qian; Dong, Ping; Lv, Xiaoyan; Cao, Yilin; Xiao, Ran

    2015-01-01

    Although the feasibility of human bone marrow mesenchymal stem cell (hBMMSC)-based tissue-engineered bone (TEB) has been proven in a number of studies, reaching a high positive fraction and bone yield of TEB still remains a challenge. Here we report a dose-effect relationship of the quantity of seeded cells with in vivo bone yield and the required quantity of hBMMSCs for the effective, stable bone formation of TEB. In our study, TEB was constructed using the static seeding technique with the gradient of seeding densities and volumes of passage 3 hBMMSCs. The in vitro characteristics of seeding efficiency, proliferation, viability, distribution, and osteogenic differentiation of hBMMSCs seeded on two commercial scaffolds of β-TCP and CHA were investigated using alamarBlue assay, live/dead staining, confocal laser scanning microscope, scanning electronic microscopy examination, and mRNA expression analysis of osteogenic differentiation markers. After 3 months of ectopic implantation, in vivo bone regeneration was examined by quantitative analysis of histology and micro-CT. The results showed that 10 × 10(6) cells/ml was the minimum cell seeding density for CHA and β-TCP to generate new bone in vivo. In addition, 20 × 10(6) cells/ml and 30 × 10(6) cells/ml were the saturating seeding densities for CHA and β-TCP to produce new bone effectively and stably, respectively. Thus, for different scaffolds, the saturating seeding density should be investigated first to ensure the effectiveness and stability of TEB construction with minimum donor injury, which is essential for the clinical application of TEB.

  11. γ-H2AX responds to DNA damage induced by long-term exposure to combined low-dose-rate neutron and γ-ray radiation.

    PubMed

    Zhang, Junlin; He, Ying; Shen, Xianrong; Jiang, Dingwen; Wang, Qingrong; Liu, Qiong; Fang, Wen

    2016-01-01

    Risk estimates for low-dose radiation (LDR) remain controversial. The possible involvement of DNA repair-related genes in long-term low-dose-rate neutron-gamma radiation exposure is poorly understood. In this study, 60 rats were divided into control groups and irradiated groups, which were exposed to low-dose-rate n-γ combined radiation (LDCR) for 15, 30, or 60 days. The effects of different cumulative radiation doses on peripheral blood cell (PBC), subsets of T cells of peripheral blood lymphocytes (PBL) and DNA damage repair were investigated. Real-time PCR and immunoblot analyses were used to detect expression of DNA DSB-repair-related genes involved in the NHEJ pathway, such as Ku70 and Ku80, in PBL. The mRNA level of H2AX and the expression level of γ-H2AX were detected by real-time PCR, immunoblot, and flow cytometry. White blood cells (WBC) and platelets (PLT) of all ionizing radiation (IR) groups decreased significantly, while no difference was seen between the 30 day and 60 day exposure groups. The numbers of CD3(+), CD4(+) T cells and CD4(+)/CD8(+) in the PBL of IR groups were lower than in the control group. In the 30 day and 60 day exposure groups, CD8(+) T cells decreased significantly. Real-time PCR and immunoblot results showed no significant difference in the mRNA and protein expression of Ku70 and Ku80 between the control groups and IR groups. However, the mRNA of H2AX increased significantly, and there was a positive correlation with dose. There was no difference in the protein expression of γ-H2AX between 30 day and 60 day groups, which may help to explain the damage to PBL. In conclusion, PBL damage increased with cumulative dose, suggesting that γ-H2AX, but neither Ku70 nor Ku80, plays an important role in PBL impairment induced by LDCR.

  12. Evaluation of the characteristics of boron-dose enhancer (BDE) materials for BNCT using near threshold 7Li(p,n)7Be direct neutrons.

    PubMed

    Bengua, Gerard; Kobayashi, Tooru; Tanaka, Kenichi; Nakagawa, Yoshinobu

    2004-03-01

    The characteristics of a number of candidate boron-dose enhancer (BDE) materials for boron neutron capture therapy (BNCT) using near threshold 7Li(p,n)7Be direct neutrons were evaluated based on the treatable protocol depth (TPD), defined in this paper. Simulation calculations were carried out by means of MCNP-4B transport code for candidate BDE materials, namely, (C2H4)n, (C2H3F)n, (C2H2F2)n, (C2HF3)n, (C2D4)n, (C2F4)n, beryllium metal, graphite, D2O and 7LiF. Dose protocols applied were those used for intra-operative BNCT treatment for brain tumour currently used in Japan. The maximum TPD (TPDmax) for each BDE material was found to be between 4 cm and 5 cm in the order of (C2H4)n < (C2H3F)n < (C2H2F2)n < (C2HF3)n < beryllium metal < (C2D4)n < graphite < (C2F4)n < D2O < 7LiF. Based on the small and arbitrary variations in the TPDmax for these materials, an explicit advantage of a candidate BDE material could not be established from the TPDmax alone. The dependence of TPD on BDE thickness was found to be influenced by the type of BDE material. For materials with hydrogen, sharp variations in TPD were observed, while those without hydrogen exhibited more moderate fluctuations in TPD as the BDE thickness was varied. The BDE thickness corresponding to TPDmax (BDE(TPDmax)) was also found to depend on the type of BDE material used. Thicker BDE(TPDmax), obtained mostly for BDE materials without hydrogen, significantly reduced the dose rates within the phantom. The TPDmax, the dependence of TPD on BDE thickness and the BDE (TPDmax) were ascertained as appropriate optimization criteria in choosing suitable BDE materials for BNCT. Among the candidate BDE materials considered in this study. (C2H4)n was judged as the suitable material for near-surface tumours and beryllium metal for deeper tumours based on these optimization criteria and other practical considerations.

  13. High-dynamic-range neutron time-of-flight detector used to infer the D(t,n)4He and D(d,n)3He reaction yield and ion temperature on OMEGA

    NASA Astrophysics Data System (ADS)

    Forrest, C. J.; Glebov, V. Yu.; Goncharov, V. N.; Knauer, J. P.; Radha, P. B.; Regan, S. P.; Romanofsky, M. H.; Sangster, T. C.; Shoup, M. J.; Stoeckl, C.

    2016-11-01

    Upgraded microchannel-plate-based photomultiplier tubes (MCP-PMT's) with increased stability to signal-shape linearity have been implemented on the 13.4-m neutron time-of-flight (nTOF) detector at the Omega Laser Facility. This diagnostic uses oxygenated xylene doped with diphenyloxazole C15H11NO + p-bis-(o-methylstyryl)-benzene (PPO + bis-MSB) wavelength shifting dyes and is coupled through four viewing ports to fast-gating MCP-PMT's, each with a different gain to allow one to measure the light output over a dynamic range of 1 × 106. With these enhancements, the 13.4-m nTOF can measure the D(t,n)4He and D(d,n)3He reaction yields and average ion temperatures in a single line of sight. Once calibrated for absolute neutron sensitivity, the nTOF detectors can be used to measure the neutron yield from 1 × 109 to 1 × 1014 and the ion temperature with an accuracy approaching 5% for both the D(t,n)4He and D(d,n)3He reactions.

  14. Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

    NASA Astrophysics Data System (ADS)

    Porollo, S. I.; Shulepin, S. V.; Konobeev, Yu. V.; Garner, F. A.

    2008-08-01

    Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.

  15. High dose neutron irradiations of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

    SciTech Connect

    Perez-Bergquist, Alex G.; Nozawa, Takashi; Shih, Chunghao Phillip; Leonard, Keith J.; Snead, Lance Lewis; Katoh, Yutai

    2014-07-01

    Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy is used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300° C. Furthermore, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.

  16. High dose neutron irradiations of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

    DOE PAGES

    Perez-Bergquist, Alex G.; Nozawa, Takashi; Shih, Chunghao Phillip; Leonard, Keith J.; Snead, Lance Lewis; Katoh, Yutai

    2014-07-01

    Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy ismore » used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300° C. Furthermore, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.« less

  17. Evaporation Residue Yields in Reactions of Heavy Neutron-Rich Radioactive Ion Beams with 64Ni and 96Zr Targets

    SciTech Connect

    Shapira, Dan; Liang, J Felix; Gross, Carl J; Varner Jr, Robert L; Beene, James R; Stracener, Daniel W; Mueller, Paul Edward; Kolata, Jim J; Roberts, Amy; Loveland, Walter; Vinodkumar, A. M.; Prisbrey, Landon; Sprunger, Peter H; Grzywacz-Jones, Kate L; Caraley, Anne L

    2009-01-01

    As hindrance sets in for the fusion of heavier systems, the effect of large neutron excess in the colliding nuclei on their probability to fuse is still an open question. The detection of evaporation residues (ERs), however, provides indisputable evidence for the fusion (complete and incomplete) in the reaction. We therefore devised a system with which we could measure ERs using low intensity neutron-rich radioactive ion beams with an efficiency close to 100%. We report on measurements of the production of ERs in collisions of {sup 132,134}Sn, {sup 134}Te and {sup 134}Sb ion beams with medium mass, neutron-rich targets. The data taken with {sup 132,134}Sn bombarding a {sup 64}Ni target are compared to available data (ERs and fusion) taken with stable Sn isotopes. Preliminary data on the fusion of {sup 132}Sn with {sup 96}Zr target are also presented.

  18. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    Thomas, David J.; Nolte, Ralf; Gressier, Vincent

    2011-12-01

    industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be

  19. Neutron- and photon-activation detection limits in breast milk analysis for prospective dose evaluation of the suckling infant.

    PubMed

    Tsipenyuk, Yu M; Firsov, V I; Cantone, M C

    2009-01-01

    Complex situations related to the environment, as in the regions affected by the Chernobyl accident and regions in which nuclear weapons testing were undertaken, as in Semipalatinsk, could be reflected in the trace element content in mothers' milk. The evaluation of fractional transfer to milk of ingested or inhaled activity and of the corresponding dose coefficients for the infant, following a mothers' radioactive intake, can take advantage from wide-ranging studies of elemental and radionuclide contents in mothers' milk. In this work the possibility to determine elements, such as Ru, Zr, Nb, Te, Ce, Th, U, in milk powder has been investigated. Although results from elemental analyses of breast milk are to be found in the literature, the determination of the identified elements has attracted poor attention since they are not considered essential elements from a biological point of view. Nevertheless, in the case of radioactive releases to the environment, such data could be of interest in evaluation of dose to the breast-fed infant. PMID:18805015

  20. Micronuclei induction in human fibroblasts exposed in vitro to Los Alamos high-energy neutrons

    NASA Astrophysics Data System (ADS)

    Gersey, Brad; Sodolak, John; Hada, Megumi; Saganti, Prem; Wilkins, Richard; Cucinotta, Francis; Wu, Honglu

    High-energy secondary neutrons, produced by the interaction of galactic cosmic rays with the atmosphere, spacecraft structure and planetary surfaces, contribute to a significant fraction to the dose equivalent in crew members and passengers during commercial aviation travel, and astronauts in space missions. The Los Alamos Nuclear Science Center (LANSCE) neutron facility's ICE House 30L beamline is known to generate neutrons that simulate the secondary neutron spectra of earth's atmosphere. The neutron spectrum is also similar to that measured onboard spacecraft like the MIR and International Space Station (ISS). To evaluate the biological damage, we exposed human fibroblasts in vitro to the LANSCE neutron beams without degrader at an entrance dose rate of 25 mGy/h and analyzed the micronuclei (MN) induction. The cells were also placed behind a 9.9 cm water column to study the effect of shielding in the protection of neutron induced damages. It was found that the dose response in the MN frequency was linear for the samples with and without shielding and the slope of the MN yield behind the shielding was reduced by a factor of 3.5. Compared to the MN induction in human fibroblasts exposed to a γ source at a similar low dose rate, the RBE was found to be 16.7 and 10.0 for the neutrons without and with the 9.9 cm water shielding, respectively.

  1. Micronuclei Induction in Human Fibroblasts Exposed In Vitro to Los Alamos High-Energy Neutrons

    NASA Technical Reports Server (NTRS)

    Gersey, Brad; Sodolak, John; Hada, Megumi; Saganti, Prem; Wilkins, Richard; Cucinotta, Francis; Wu, Honglu

    2006-01-01

    High-energy secondary neutrons, produced by the interaction of galactic cosmic rays with the atmosphere, spacecraft structure and planetary surfaces, contribute to a significant fraction to the dose equivalent in crew members and passengers during commercial aviation travel, and astronauts in space missions. The Los Alamos Nuclear Science Center (LANSCE) neutron facility#s ICE House 30L beamline is known to generate neutrons that simulate the secondary neutron spectra of earth#s atmosphere. The neutron spectrum is also similar to that measured onboard spacecraft like the MIR and International Space Station (ISS). To evaluate the biological damage, we exposed human fibroblasts in vitro to the LANSCE neutron beams without degrader at an entrance dose rate of 25 mGy/hr and analyzed the micronuclei (MN) induction. The cells were also placed behind a 9.9 cm water column to study effect of shielding in the protection of neutron induced damages. It was found that the dose response in the MN frequency was linear for the samples with and without shielding and the slope of the MN yield behind the shielding was reduced by a factor of 3.5. Compared to the MN induction in human fibroblasts exposed to a gamma source at a low dose rate, the RBE was found to be 16.7 and 10.0 for the neutrons without and with 9.9 cm water shielding, respectively.

  2. Energy Dependence of Fission Product Yields from 235U, 238U and 239Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, M.; Arnold, C.; Bredeweg, T.; Vieira, D.; Wilhelmy, J.; Tonchev, A.; Stoyer, M.; Bhike, M.; Krishichayan, F.; Tornow, W.; Fowler, M.

    2015-10-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements has been performed. The energy dependence of a number of cumulative fission product yields (FPY) have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of fission counting using specially designed dual-fission chambers and ?-ray counting. Each dual-fission chamber is a back-to-back ionization chamber encasing an activation target in the center with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the total number of fissions in the activation target with no reference to the fission cross-section, thus reducing uncertainties. ?-ray counting of the activation target was performed on well-shielded HPGe detectors over a period of 2 months post irradiation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 3.6, 4.6, 5.5, 7.5, 8.9 and 14.8 MeV. These results are compared to previous measurements and theoretical estimates. This work was performed under the auspices of the USDoE by Los Alamos National Security, LLC under Contract DE-AC52-06NA25396.

  3. Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

    2009-06-01

    At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

  4. Human circulating plasma DNA significantly decreases while lymphocyte DNA damage increases under chronic occupational exposure to low-dose gamma-neutron and tritium β-radiation.

    PubMed

    Korzeneva, Inna B; Kostuyk, Svetlana V; Ershova, Liza S; Osipov, Andrian N; Zhuravleva, Veronika F; Pankratova, Galina V; Porokhovnik, Lev N; Veiko, Natalia N

    2015-09-01

    The blood plasma of healthy people contains cell-fee (circulating) DNA (cfDNA). Apoptotic cells are the main source of the cfDNA. The cfDNA concentration increases in case of the organism's cell death rate increase, for example in case of exposure to high-dose ionizing radiation (IR). The objects of the present research are the blood plasma and blood lymphocytes of people, who contacted occupationally with the sources of external gamma/neutron radiation or internal β-radiation of tritium N = 176). As the controls (references), blood samples of people, who had never been occupationally subjected to the IR sources, were used (N = 109). With respect to the plasma samples of each donor there were defined: the cfDNA concentration (the cfDNA index), DNase1 activity (the DNase1 index) and titre of antibodies to DNA (the Ab DNA index). The general DNA damage in the cells was defined (using the Comet assay, the tail moment (TM) index). A chronic effect of the low-dose ionizing radiation on a human being is accompanied by the enhancement of the DNA damage in lymphocytes along with a considerable cfDNA content reduction, while the DNase1 content and concentration of antibodies to DNA (Ab DNA) increase. All the aforementioned changes were also observed in people, who had not worked with the IR sources for more than a year. The ratio cfDNA/(DNase1×Ab DNA × TM) is proposed to be used as a marker of the chronic exposure of a person to the external low-dose IR. It was formulated the assumption that the joint analysis of the cfDNA, DNase1, Ab DNA and TM values may provide the information about the human organism's cell resistivity to chronic exposure to the low-dose IR and about the development of the adaptive response in the organism that is aimed, firstly, at the effective cfDNA elimination from the blood circulation, and, secondly - at survival of the cells, including the cells with the damaged DNA. PMID:26113293

  5. Biological effectiveness of neutron irradiation on animals and man

    SciTech Connect

    Straume, T.

    1982-11-01

    Neutron experiments on a highly radiosensitive in vivo system - oocytes in mice - provide new insight into the nature of the radiosensitive targets of these important cells. With the radiobiological literature as background, neutron data from animals and humans are integrated, and the controversial question of radiation protection standards for neutrons is addressed. Oocyte killing in juvenile mice by 0.43-MeV, /sup 252/Cf-fission, and 15 MeV neutrons, compared with that by /sup 60/Co gamma rays, yields unusually low neutron RBEs (relative biological effectiveness). At 0.1 rad of 0.43-MeV neutrons the RBE is only 1.8, contrasting greatly with values of 100 or more reported at low-doses for other endpoints. In mice just prior to birth, however, when oocytes are less radiosensitive, the neutron RBE is much higher, similar to values for most other mammalian endpoints. This dramatic change in neutron RBE with mouse age (occurring within 2 to 3 days) can be explained as the result of a shift from a less radiosensitive target (presumably nuclear DNA) to a much more radiosensitive one (probably the oocyte plasma membrane). Using various approaches, a value for the neutron Quality Factor (Q, a radiation protection standard) is estimated as 17 (+-100%), much lower than 100 which has been suggested. With the large uncertainty, 17 is not markedly different from the value of 10 presently in general use.

  6. Predicted yields of new neutron-rich isotopes of nuclei with Z=64-80 in the multinucleon transfer reaction {sup 48}Ca+{sup 238}U

    SciTech Connect

    Adamian, G. G.; Antonenko, N. V.; Sargsyan, V. V.; Scheid, W.

    2010-05-15

    The production cross sections of new neutron-rich isotopes of nuclei with charge numbers Z=64-80 are estimated for future experiments in the multinucleon transfer reaction {sup 48}Ca+{sup 238}U at bombarding energy E{sub c.m.}=189 MeV close to the Coulomb barrier.

  7. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    Thomas, David J.; Nolte, Ralf; Gressier, Vincent

    2011-12-01

    industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be

  8. Accelerator shield design of KIPT neutron source facility

    SciTech Connect

    Zhong, Z.; Gohar, Y.

    2013-07-01

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

  9. Portable neutron spectrometer and dosimeter

    DOEpatents

    Waechter, David A.; Erkkila, Bruce H.; Vasilik, Dennis G.

    1985-01-01

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  10. Portable neutron spectrometer and dosimeter

    DOEpatents

    Waechter, D.A.; Erkkila, B.H.; Vasilik, D.G.

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  11. MONDO: A neutron tracker for particle therapy secondary emission fluxes measurements

    NASA Astrophysics Data System (ADS)

    Marafini, M.; Patera, V.; Pinci, D.; Sarti, A.; Sciubba, A.; Spiriti, E.

    2016-07-01

    Cancer treatment is performed, in Particle Therapy, using accelerated charged particles whose high irradiation precision and conformity allows the tumor destruction while sparing the surrounding healthy tissues. Dose release monitoring devices using photons and charged particles produced by the beam interaction with the patient body have already been proposed, but no attempt based on the detection of the abundant secondary radiation neutron component has been made yet. The reduced attenuation length of neutrons yields a secondary particle sample that is larger in number when compared to photons and charged particles. Furthermore, neutrons allow for a backtracking of the emission point that is not affected by multiple scattering. Since neutrons can release a significant dose far away from the tumor region, a precise measurement of their flux, production energy and angle distributions is eagerly needed in order to improve the Treatment Planning Systems (TPS) software, so to predict not only the normal tissue toxicity in the target region but also the risk of late complications in the whole body. All the aforementioned issues underline the importance for an experimental effort devoted to the precise characterization of the neutron production gaining experimental access both to the emission point and production energy. The technical challenges posed by a neutron detector aiming for a high detection efficiency and good backtracking precision will be addressed within the MONDO (MOnitor for Neutron Dose in hadrOntherapy) project. The MONDO's main goal is to develop a tracking detector targeting fast and ultrafast secondary neutrons. The tracker is composed by a scintillating fiber matrix (4 × 4 × 8cm3). The full reconstruction of protons, produced in elastic interactions, will be used to measure energy and direction of the impinging neutron. The neutron tracker will measure the neutron production yields, as a function of production angle and energy, using different

  12. Neutron Skyshine Considerations For The NIF Shielding Design

    SciTech Connect

    Singh, M S; Mecozzi, J M; Tobin, M T

    2004-01-28

    A series of coupled neutron-photon transport Monte-Carlo calculations was performed to estimate the roof shielding required to limit the skyshine dose to less than 1 mrem/y at the site boundary when conducting DT experiments with annual fusion yields up to 1200 MJ (4.2E20 neutrons/y). The NIF shielding design consists of many different components. The basic components include 10-cm-thick Al chamber with 40-cm-thick target chamber gunite shield having multiple penetrations, 1.83-m-thick concrete Target Bay walls, 1.37-m-thick concrete roof, and multiple concrete floors with numerous penetrations. Under this shielding configuration, the skyshine dose at the nearest site-boundary was calculated to be less than 0.2 mrem/y for all possible target illumination configurations. The potential dose at the site boundary would be about one-tenth of the cosmic neutron dose that we measured with bubble neutron detectors on board a commercial roundtrip flight from SF to Rochester. This incremental dose increase is well within the normal fluctuations (noise) of the natural background radiation in the Livermore area. The skyshine dose has no impact on the public. The skyshine dose trends at ground and elevated levels are plotted as a function of distance from 20 m to 1000 m from the center of the target bay. The differential neutron and photon energy flux emerging from the NIF roof and at several locations on the ground is plotted to show how it shifts with distance. The results of this study are compared with the neutron skyshine studies done at high-energy accelerators by R. H. Thomas.

  13. The g13 Experiment at Jefferson Lab: Strangeness Production on the Neutron in the Deuteron with Polarized Photons: {gamma}-vectorn{yields}KY-vector

    SciTech Connect

    Munevar, E.; Berman, B. L.; Nadel-Turonski, P.

    2007-10-26

    Strangeness has been shown to be important for the understanding of the so-called missing resonances. Due to the scarce experimental data in strangeness photoproduction on the neutron, phenomenological models such as coupled-channels analyses resort to certain approximations that do not allow getting either accuracy or agreement between different analyses when extracting resonance parameters. Thus, in order to obtain high-quality data on the neutron channels, a new experiment (designated g13), based on a liquid deuterium target and a polarized photon beam (both circular and linear polarization) covering from threshold to 2.3 GeV has been done at the Thomas Jefferson National Accelerator Facility. In this paper, a brief description of the g13 experiment is given.

  14. Energy Dependence of Fission Product Yields from 235U, 238U and 239Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, M. E.; Arnold, C. W.; Becker, J. A.; Bhatia, C.; Bhike, M.; Bond, E. M.; Bredeweg, T. A.; Fallin, B.; Fowler, M. M.; Howell, C. R.; Kelley, J. H.; Krishichayan; Macri, R.; Rusev, G.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.; Tornow, W.; Vieira, D. J.; Wilhelmy, J. B.

    2016-01-01

    Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varying degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for 235U, 238U and 239Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission chamber and gamma

  15. Energy dependence of fission product yields from 235U, 238U and 239Pu for incident neutron energies between 0.5 and 14.8 MeV

    DOE PAGES

    Gooden, M. E.; Arnold, C. W.; Becker, J. A.; Bhatia, C.; Bhike, M.; Bond, E. M.; Bredeweg, T. A.; Fallin, B.; Fowler, M. M.; Howell, C. R.; et al

    2016-01-06

    In this study, Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varyingmore » degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for 235U, 238U and 239Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual

  16. L-Phenylalanine preloading reduces the (10)B(n, α)(7)Li dose to the normal brain by inhibiting the uptake of boronophenylalanine in boron neutron capture therapy for brain tumours.

    PubMed

    Watanabe, Tsubasa; Tanaka, Hiroki; Fukutani, Satoshi; Suzuki, Minoru; Hiraoka, Masahiro; Ono, Koji

    2016-01-01

    Boron neutron capture therapy (BNCT) is a cellular-level particle radiation therapy that combines the selective delivery of boron compounds to tumour tissue with neutron irradiation. Previously, high doses of one of the boron compounds used for BNCT, L-BPA, were found to reduce the boron-derived irradiation dose to the central nervous system. However, injection with a high dose of L-BPA is not feasible in clinical settings. We aimed to find an alternative method to improve the therapeutic efficacy of this therapy. We examined the effects of oral preloading with various analogues of L-BPA in a xenograft tumour model and found that high-dose L-phenylalanine reduced the accumulation of L-BPA in the normal brain relative to tumour tissue. As a result, the maximum irradiation dose in the normal brain was 19.2% lower in the L-phenylalanine group relative to the control group. This study provides a simple strategy to improve the therapeutic efficacy of conventional boron compounds for BNCT for brain tumours and the possibility to widen the indication of BNCT to various kinds of other tumours. PMID:26455769

  17. L-Phenylalanine preloading reduces the (10)B(n, α)(7)Li dose to the normal brain by inhibiting the uptake of boronophenylalanine in boron neutron capture therapy for brain tumours.

    PubMed

    Watanabe, Tsubasa; Tanaka, Hiroki; Fukutani, Satoshi; Suzuki, Minoru; Hiraoka, Masahiro; Ono, Koji

    2016-01-01

    Boron neutron capture therapy (BNCT) is a cellular-level particle radiation therapy that combines the selective delivery of boron compounds to tumour tissue with neutron irradiation. Previously, high doses of one of the boron compounds used for BNCT, L-BPA, were found to reduce the boron-derived irradiation dose to the central nervous system. However, injection with a high dose of L-BPA is not feasible in clinical settings. We aimed to find an alternative method to improve the therapeutic efficacy of this therapy. We examined the effects of oral preloading with various analogues of L-BPA in a xenograft tumour model and found that high-dose L-phenylalanine reduced the accumulation of L-BPA in the normal brain relative to tumour tissue. As a result, the maximum irradiation dose in the normal brain was 19.2% lower in the L-phenylalanine group relative to the control group. This study provides a simple strategy to improve the therapeutic efficacy of conventional boron compounds for BNCT for brain tumours and the possibility to widen the indication of BNCT to various kinds of other tumours.

  18. Nuclear and dosimetric features of an isotopic neutron source

    NASA Astrophysics Data System (ADS)

    Vega-Carrillo, H. R.; Hernández-Dávila, V. M.; Rivera, T.; Sánchez, A.

    2014-02-01

    A multisphere neutron spectrometer was used to determine the features of a 239PuBe neutron source that is used to operate the ESFM-IPN Subcritical Reactor. To determine the source main features it was located a 100 cm from the spectrometer which was a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter polyethylene spheres. Count rates obtained with the spectrometer were unfolded using the NSDUAZ code and neutron spectrum, total fluence, and ambient dose equivalent were determined. A Monte Carlo calculation was carried out to estimate the spectrum and integral features being less than values obtained experimentally due to the presence of 241Pu in the Pu used to fabricate the source. Actual neutron yield and the mass fraction of 241Pu was estimated.

  19. Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air.

    1997-06-09

    Version 01 SKYSHINE was designed to aid in the evaluation of the effects of structure geometry on the gamma-ray dose rate at given detector positions outside of a building housing N16 gamma-ray sources. The program considers a rectangular structure enclosed by four walls and a roof. Each of the walls and the roof of the building may be subdivided into up to nine different areas, representing different materials or different thicknesses of the same materialmore » for those positions of the wall or roof. Basic sets of iron and concrete slab transmission and reflection data for 6.2 MeV gamma rays are part of the SKYSHINE block data. These data, as well as parametric air transport data for line-beam sources at a number of energies between 0.6 MeV and 6.2 MeV and ranges to 3750 ft, are used to estimate the various components of the gamma-ray dose rate at positions outside of the building. The gamma-ray source is assumed to be a 6.2-MeV point-isotropic source. SKYSHINE-III provides an increase in versatility over the original SKYSHINE code in that it addresses both neutron and gamma-ray point sources. In addition, the emitted radiation may be characterized by an energy emission spectrum defined by the user. A new SKYSHINE data base is also included. SKYIII-PC is a PC version of SKYSHINE-III. Only minor modifications were made in converting for PC use. The June 1997 replacement of the PC version corrects the previously existing index problem leading to erroneous results for the "wall-scattered/air-scattered" contribution if a roof is modeled. Associated with these changes is the precaution that the detector height should always be lower than the base of the roof. Erroneous results for the roof portion of the "wall-scattered/air- attenuated" contribution will occur if a roof is modeled and the detector is not below the roof plane.« less

  20. Numerical model for computation of effective and ambient dose equivalent at flight altitudes. Application for dose assessment during GLEs

    NASA Astrophysics Data System (ADS)

    Mishev, Alexander; Usoskin, Ilya

    2015-05-01

    A numerical model for assessment of the effective dose and ambient dose equivalent produced by secondary cosmic ray particles of galactic and solar origin at commercial aircraft altitudes is presented. The model represents a full chain analysis based on ground-based measurements of cosmic rays, from particle spectral and angular characteristics to dose estimation. The model is based on newly numerically computed yield functions and realistic propagation of cosmic ray in the Earth magnetosphere. The yield functions are computed using a straightforward full Monte Carlo simulation of the atmospheric cascade induced by primary protons and α-particles and subsequent conversion of secondary particle fluence (neutrons, protons, gammas, electrons, positrons, muons and charged pions) to effective dose or the ambient dose equivalent. The ambient dose equivalent is compared with reference data at various conditions such as rigidity cut-off and level of solar activity. The method is applied for computation of the effective dose rate at flight altitude during the ground level enhancement of 13 December 2006. The solar proton spectra are derived using neutron monitor data. The computation of the effective dose rate during the event explicitly considers the derived anisotropy i.e. the pitch angle distribution as well as the propagation of the solar protons in the magnetosphere of the Earth.

  1. High-Dose Sirolimus And Immune Selective Pentostatin Plus Cyclophosphamide Conditioning Yields Stable Mixed Chimerism and Insufficient Graft-Versus-Tumor Responses

    PubMed Central

    Mossoba, Miriam E.; Halverson, David C.; Kurlander, Roger; Schuver, Bazetta Blacklock; Carpenter, Ashley; Hansen, Brenna; Steinberg, Seth M.; Ali, Syed Abbas; Tageja, Nishant; Hakim, Frances T.; Gea-Banacloche, Juan; Sportes, Claude; Hardy, Nancy M.; Hickstein, Dennis D.; Pavletic, Steven Z.; Khuu, Hanh; Sabatini, Marianna; Stroncek, David; Levine, Bruce L.; June, Carl H.; Mariotti, Jacopo; Rixe, Olivier; Fojo, Antonio Tito; Bishop, Michael R.; Gress, Ronald. E.; Fowler, Daniel H.

    2015-01-01

    PURPOSE We hypothesized that lymphoid-selective host conditioning and subsequent adoptive transfer of sirolimus-resistant allogeneic T cells (T-Rapa), when combined with high-dose sirolimus drug therapy in vivo, would safely achieve anti-tumor effects while avoiding GVHD. EXPERIMENTAL DESIGN Patients (n=10) with metastatic renal cell carcinoma (RCC) were accrued because this disease is relatively refractory to high-dose conditioning yet may respond to high-dose sirolimus. A 21-day outpatient regimen of weekly pentostatin (P; 4 mg/m2/dose) combined with daily, dose-adjusted cyclophosphamide (C; ≤ 200 mg/day) was designed to deplete and suppress host T cells. After PC conditioning, patients received matched sibling, T cell-replete peripheral blood stem cell allografts and high-dose sirolimus (serum trough target, 20–30 ng/ml). To augment graft-versus-tumor (GVT) effects, multiple T-Rapa donor lymphocyte infusions (DLI) were administered (days 0, 14, and 45 post-transplant) and sirolimus was discontinued early (day 60 post-transplant). RESULTS PC conditioning depleted host T cells without neutropenia or infection and facilitated donor engraftment (10/10 cases). High-dose sirolimus therapy inhibited multiple T-Rapa DLI, as evidenced by stable mixed donor/host chimerism. No anti-tumor responses were detected by RECIST criteria and no significant classical acute GVHD was observed. CONCLUSIONS Immune-selective PC conditioning represents a new approach to safely achieve alloengraftment without neutropenia. However, allogeneic T cells generated ex vivo in sirolimus are not resistant to the tolerance-inducing effects of in vivo sirolimus drug therapy, thereby cautioning against use of this intervention in patients with refractory cancer. PMID:26071480

  2. POST-SHOT RADIATION ENVIRONMENT FOLLOWING LOW-YIELD SHOTS INSIDE THE NATIONAL IGNITION FACILITY

    SciTech Connect

    Sitaraman, S; Brereton, S; Dauffy, L; Hall, J; Hansen, L; Khater, H; Kim, S; Pohl, B; Verbeke, J

    2010-10-29

    A detailed model of the Target Bay (TB) at the National Ignition Facility (NIF) has been developed to estimate the post-shot radiation environment inside the facility. The model includes large number of structures and diagnostic instruments present inside the TB. These structures and instruments are activated by the few nanosecond pulse of neutrons generated during a shot and the resultant gamma dose rates are estimated at various decay times following the shot. The results presented in this paper are based on a low-yield D-T shot of 10{sup 16} neutrons. General environment dose rates drop to below 3 mrem/h within three hours following a shot with higher dose rates observed at contact with some of the components. Dose rate maps of the different TB levels were generated to aid in estimating worker stay-out times following a shot before entry is permitted into the TB.

  3. Improvement effect on the depth-dose distribution by CSF drainage and air infusion of a tumour-removed cavity in boron neutron capture therapy for malignant brain tumours.

    PubMed

    Sakurai, Yoshinori; Ono, Koji; Miyatake, Shin-Ichi; Maruhashi, Akira

    2006-03-01

    Boron neutron capture therapy (BNCT) without craniotomy for malignant brain tumours was started using an epi-thermal neutron beam at the Kyoto University Reactor in June 2002. We have tried some techniques to overcome the treatable-depth limit in BNCT. One of the effective techniques is void formation utilizing a tumour-removed cavity. The tumorous part is removed by craniotomy about 1 week before a BNCT treatment in our protocol. Just before the BNCT irradiation, the cerebro-spinal fluid (CSF) in the tumour-removed cavity is drained out, air is infused to the cavity and then the void is made. This void improves the neutron penetration, and the thermal neutron flux at depth increases. The phantom experiments and survey simulations modelling the CSF drainage and air infusion of the tumour-removed cavity were performed for the size and shape of the void. The advantage of the CSF drainage and air infusion is confirmed for the improvement in the depth-dose distribution. From the parametric surveys, it was confirmed that the cavity volume had good correlation with the improvement effect, and the larger effect was expected as the cavity volume was larger.

  4. Characterisation of neutron fields at Cernavoda NPP.

    PubMed

    Cauwels, Vanessa; Vanhavere, Filip; Dumitrescu, Dorin; Chirosca, Alecsandru; Hager, Luke; Million, Marc; Bartz, James

    2013-04-01

    Near a nuclear reactor or a fuel container, mixed neutron/gamma fields are very common, necessitating routine neutron dosimetry. Accurate neutron dosimetry is complicated by the fact that the neutron effective dose is strongly dependent on the neutron energy and the direction distribution of the neutron fluence. Neutron field characterisation is indispensable if one wants to obtain a reliable estimate for the neutron dose. A measurement campaign at CANDU nuclear power plant located in Cernavoda, Romania, was set up to characterise the neutron fields in four different locations and to investigate the behaviour of different neutron personal dosemeters. This investigation intends to assist in choosing a suitable neutron dosimetry system at this nuclear power plant.

  5. Radiological dose assessment for vault storage concepts

    SciTech Connect

    Richard, R.F.

    1997-02-25

    This radiological dose assessment presents neutron and photon dose rates in support of project W-460. Dose rates are provided for a single 3013 container, the ``infloor`` storage vault concept, and the ``cubicle`` storage vault concept.

  6. Gadolinium neutron capture brachytherapy (GdNCB), a new treatment method for intravascular brachytherapy

    SciTech Connect

    Enger, Shirin A.; Rezaei, Arash; Munck af Rosenschoeld, Per; Lundqvist, Hans

    2006-01-15

    Restenosis is a major problem after balloon angioplasty and stent implantation. The aim of this study is to introduce gadolinium neutron capture brachytherapy (GdNCB) as a suitable modality for treatment of stenosis. The utility of GdNCB in intravascular brachytherapy (IVBT) of stent stenosis is investigated by using the GEANT4 and MCNP4B Monte Carlo radiation transport codes. To study capture rate, Kerma, absorbed dose and absorbed dose rate around a Gd-containing stent activated with neutrons, a 30 mm long, 5 mm diameter gadolinium foil is chosen. The input data is a neutron spectrum used for clinical neutron capture therapy in Studsvik, Sweden. Thermal neutron capture in gadolinium yields a spectrum of high-energy gamma photons, which due to the build-up effect gives an almost flat dose delivery pattern to the first 4 mm around the stent. The absorbed dose rate is 1.33 Gy/min, 0.25 mm from the stent surface while the dose to normal tissue is in order of 0.22 Gy/min, i.e., a factor of 6 lower. To spare normal tissue further fractionation of the dose is also possible. The capture rate is relatively high at both ends of the foil. The dose distribution from gamma and charge particle radiation at the edges and inside the stent contributes to a nonuniform dose distribution. This will lead to higher doses to the surrounding tissue and may prevent stent edge and in-stent restenosis. The position of the stent can be verified and corrected by the treatment plan prior to activation. Activation of the stent by an external neutron field can be performed days after catherization when the target cells start to proliferate and can be expected to be more radiation sensitive. Another advantage of the nonradioactive gadolinium stent is the possibility to avoid radiation hazard to personnel.

  7. PFP vertical calciner shield wall dose rate calculations using MCNP

    SciTech Connect

    Wittekind, W.D.

    1997-08-21

    This report yields a neutron shield wall design for a full time occupancy dose rate of 0.25 mrem/h. ORIGEN2 generated gamma ray spectrum and neutron intensity for plutonium. MCNP modeled the calciner glovebox and room for reflection of neutrons off concrete walls and ceiling. Neutron calculations used MCNP in mode n, p to include neutron capture gammas. Photon calculations used MCNP in mode p for gamma rays. Neutron shield with lower 137.16 cm (4.5 feet) of 12.7 cm (5 inch) thick Lucite{reg_sign} and 0.3175 cm (0.125 inch) stainless steel on both sides, and upper 76.2 cm (2.5 feet) of 10.16 cm (4 inch) thick Lucite{reg_sign} and 1.905 cm (0.75 inch) thick glass on each side gave a total weighted dose rate of 0.23 mrem/h, fulfilling the design goal. Lucite{reg_sign} is considered to be equivalent to Plexiglas{reg_sign} since both are methylmethacrylate polymers.

  8. Pulsed neutron detector

    DOEpatents

    Robertson, deceased, J. Craig; Rowland, Mark S.

    1989-03-21

    A pulsed neutron detector and system for detecting low intensity fast neutron pulses has a body of beryllium adjacent a body of hydrogenous material the latter of which acts as a beta particle detector, scintillator, and moderator. The fast neutrons (defined as having En>1.5 MeV) react in the beryllium and the hydrogenous material to produce larger numbers of slow neutrons than would be generated in the beryllium itself and which in the beryllium generate hellium-6 which decays and yields beta particles. The beta particles reach the hydrogenous material which scintillates to yield light of intensity related to the number of fast neutrons. A photomultiplier adjacent the hydrogenous material (scintillator) senses the light emission from the scintillator. Utilization means, such as a summing device, sums the pulses from the photo-multiplier for monitoring or other purposes.

  9. Thermal neutron response of the Li 2B 4O 7:Cu TL dosimeter

    NASA Astrophysics Data System (ADS)

    Gauld, I. C.; Harvey, J. W.; Kennett, T. J.; Prestwich, W. V.

    1986-10-01

    A measurement of the thermal neutron response of the Li 2B 4O 7:Cu TLD has been conducted. The results obtained using the Panasonic UD-806 dosimeter and UD-854A holder yield a free-in-air response of 3.3±0.1 R 60Co equivalent per mSv of thermal neutrons. A thermal neutron response of over 7 R 60Co equivalent per mSv was observed when the dosimeter was irradiated on water phantoms. The high sensitivity may result in a substantial overestimate of the gamma dose equivalent if the TLD is used in a mixed neutron and gamma environment of unknown ratio. Measurements of the Li 2B 4O 7:Cu glow curve, TL saturation curve and the thermal neutron response dependence of the dosimeter filtration thickness are also presented.

  10. Liposomal Nanoparticles of a Spleen Tyrosine Kinase P-Site Inhibitor Amplify the Potency of Low Dose Total Body Irradiation Against Aggressive B-Precursor Leukemia and Yield Superior Survival Outcomes in Mice.

    PubMed

    Uckun, Fatih M; Myers, Dorothea E; Cheng, Jianjun; Qazi, Sanjive

    2015-06-01

    This study was designed to improve the efficacy of radiation therapy against radiation-resistant leukemia. We report that the potency of low dose radiation therapy against B-precursor acute lymphoblastic leukemia (BPL) can be markedly enhanced by combining radiation with a liposomal nanoparticle (LNP) formulation of the SYK-P-site inhibitor C61 ("C61-LNP"). C61-LNP plus low dose total body irradiation (TBI) was substantially more effective than TBI alone or C61-LNP alone in improving the event-free survival outcome NOD/SCID mice challenged with an otherwise invariably fatal dose of human ALL xenograft cells derived from relapsed BPL patients. C61-LNP plus low dose TBI also yielded progression-free survival, tumor-free survival and overall survival outcomes in CD22ΔE12 × BCR-ABL double transgenic mice with advanced stage, radiation-resistant BPL with lymphomatous features that were significantly superior to those of mice treated with TBI alone or C61-LNP alone.

  11. Boron neutron capture therapy (BNCT) for liver metastasis in an experimental model: dose–response at five-week follow-up based on retrospective dose assessment in individual rats

    SciTech Connect

    Emiliano C. C. Pozzi; Veronica A. Trivilin; Lucas L. Colombo; Andrea Monti Hughes; Silvia I. Thorp; Jorge E. Cardoso; Marcel A. Garabalino; Ana J. Molinari; Elisa M. Heber; Paula Curotto; Marcelo Miller; Maria E. Itoiz; Romina F. Aromando; David W. Nigg; Amanda E. Schwint

    2013-11-01

    Boron neutron capture therapy (BNCT) was proposed for untreatable colorectal liver metastases. Employing an experimental model of liver metastases in rats, we recently demonstrated that BNCT mediated by boronophenylalanine (BPA-BNCT) at 13 Gy prescribed to tumor is therapeutically useful at 3-week follow-up. The aim of the present study was to evaluate dose–response at 5-week follow-up, based on retrospective dose assessment in individual rats. BDIX rats were inoculated with syngeneic colon cancer cells DHD/K12/TRb. Tumor-bearing animals were divided into three groups: BPA-BNCT (n = 19), Beam only (n = 8) and Sham (n = 7) (matched manipulation, no treatment). For each rat, neutron flux was measured in situ and boron content was measured in a pre-irradiation blood sample for retrospective individual dose assessment. For statistical analysis (ANOVA), individual data for the BPA-BNCT group were pooled according to absorbed tumor dose, BPA-BNCT I: 4.5–8.9 Gy and BPA-BNCT II: 9.2–16 Gy. At 5 weeks post-irradiation, the tumor surface area post-treatment/pre-treatment ratio was 12.2 +/- 6.6 for Sham, 7.8 +/- 4.1 for Beam only, 4.4 +/- 5.6 for BPA-BNCT I and 0.45 +/- 0.20 for BPA-BNCT II; tumor nodule weight was 750 +/- 480 mg for Sham, 960 +/- 620 mg for Beam only, 380 +/- 720 mg for BPA-BNCT I and 7.3 +/- 5.9 mg for BPA-BNCT II. The BPA-BNCT II group exhibited statistically significant tumor control with no contributory liver toxicity. Potential threshold doses for tumor response and significant tumor control were established at 6.1 and 9.2 Gy, respectively.

  12. Temperature dependence of the deformation behavior of 316 stainless steel after low temperature neutron irradiation

    SciTech Connect

    Pawel-Robertson, J.E.; Rowcliffe, A.F.; Grossbeck, M.L.

    1996-10-01

    The effects of low temperature neutron irradiation on the tensile behavior of 316 stainless steel have been investigated. A single heat of solution annealed 316 was irradiated to 7 and 18 dpa at 60, 200, 330, and 400{degrees}C. The tensile properties as a function of dose and as a function of temperature were examined. Large changes in yield strength, deformation mode, strain to necking, and strain hardening capacity were seen in this irradiation experiment. The magnitudes of the changes are dependent on both irradiation temperature and neutron dose. Irradiation can more than triple the yield strength over the unirradiated value and decrease the strain to necking (STN) to less than 0.5% under certain conditions. A maximum increase in yield strength and a minimum in the STN occur after irradiation at 330{degrees}C but the failure mode remains ductile.